WorldWideScience

Sample records for es-salam research reactor

  1. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  2. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  3. Reactor Safety Research Programs

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  4. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  5. Gaseous fuel reactor research

    Science.gov (United States)

    Thom, K.; Schneider, R. T.

    1977-01-01

    The paper reviews studies dealing with the concept of a gaseous fuel reactor and describes the structure and plans of the current NASA research program of experiments on uranium hexafluoride systems and uranium plasma systems. Results of research into the basic properties of uranium plasmas and fissioning gases are reported. The nuclear pumped laser is described, and the main results of experiments with these devices are summarized.

  6. CER. Research reactors in France

    Energy Technology Data Exchange (ETDEWEB)

    Estrade, Jerome [CEA, DEN, DER, Saint-Paul-lez-Durance (France). Jules Horowitz Reactor (JHR)

    2012-10-15

    Networking and the establishment of coalitions between research reactors are important to guarantee a high technical quality of the facility, to assure well educated and trained personnel, to harmonize the codes of standards and the know-ledge of the personnel as well as to enhance research reactor utilization. In addition to the European co-operation, country-specific working groups have been established for many years, such as the French research reactor Club d'Exploitants des Reacteurs (CER). It is the association of French research reactors representing all types of research reactors from zero power up to high flux reactors. CER was founded in 1990 and today a number of 14 research reactors meet twice a year for an exchange of experience. (orig.)

  7. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  8. Utilisation of British University Research Reactors.

    Science.gov (United States)

    Duncton, P. J.; And Others

    British experience relating to the employment of university research reactors and subcritical assemblies in the education of nuclear scientists and technologists, in the training of reactor operators and for fundamental pure and applied research in this field is reviewed. The facilities available in a number of British universities and the uses…

  9. Performance of a multipurpose research electrochemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Henquin, E.R. [Programa de Electroquimica Aplicada e Ingenieria Electroquimica (PRELINE), Facultad de Ingenieria Quimica, Universidad Nacional del Litoral, Santiago del Estero 2829, S3000AOM Santa Fe (Argentina); Bisang, J.M., E-mail: jbisang@fiq.unl.edu.ar [Programa de Electroquimica Aplicada e Ingenieria Electroquimica (PRELINE), Facultad de Ingenieria Quimica, Universidad Nacional del Litoral, Santiago del Estero 2829, S3000AOM Santa Fe (Argentina)

    2011-07-01

    Highlights: > For this reactor configuration the current distribution is uniform. > For this reactor configuration with bipolar connection the leakage current is small. > The mass-transfer conditions are closely uniform along the electrode. > The fluidodynamic behaviour can be represented by the dispersion model. > This reactor represents a suitable device for laboratory trials. - Abstract: This paper reports on a multipurpose research electrochemical reactor with an innovative design feature, which is based on a filter press arrangement with inclined segmented electrodes and under a modular assembly. Under bipolar connection, the fraction of leakage current is lower than 4%, depending on the bipolar Wagner number, and the current distribution is closely uniform. When a turbulence promoter is used, the local mass-transfer coefficient shows a variation of {+-}10% with respect to its mean value. The fluidodynamics of the reactor responds to the dispersion model with a Peclet number higher than 10. It is concluded that this reactor is convenient for laboratory research.

  10. Education and Training on ISIS Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X. [French Atomic Energy and Alternative Energies Commission, Paris (France)

    2013-07-01

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions.

  11. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

    2000-05-01

    The fabrication technology of the U{sub 3}Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U{sub 3}Si{sub 2} dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U{sub 3}Si{sub 2} fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 {approx} 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The {gamma}-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U{sub 3}Si{sub 2}. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano

  12. Reactor containment research and development

    Energy Technology Data Exchange (ETDEWEB)

    Weil, N. A.

    1963-06-15

    An outline is given of containment concepts, sources and release rates of energy, responses of containment structures, effects of projectiles, and leakage rates of radioisotopes, with particular regard to major reactor accidents. (T.F.H.)

  13. Gaseous fuel nuclear reactor research

    Science.gov (United States)

    Schwenk, F. C.; Thom, K.

    1975-01-01

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

  14. TRIGA research reactors; Reacteurs de recherche Triga

    Energy Technology Data Exchange (ETDEWEB)

    Fouquet, D.M.; Razvi, J.; Whittemore, W.L. [Triga General Atomics, San Diego, CA (United States); Duban, B.; Harbonnier, G.; Du Limbert, P.; Durand, J.P. [AREVA/FRAMATOME ANP/CERCA, 92 - Paris-La-Defence (France)

    2004-02-01

    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.)

  15. New research possibilities at the Budapest research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hargitai, T.; Vidovszky, I. [KFKI Atomic Energy Research Institute, H-1525 Budapest (Hungary)

    2001-07-01

    The Budapest Research Reactor is the first nuclear facility of Hungary. It was commissioned in 1959, reconstructed and upgraded in 1967 and 1986-92. The main purpose of the reactor is to serve neutron research. The reactor was extended by a liquid hydrogen type cold neutron source in 2000. The research possibilities are much improved by the CNS both in neutron scattering and neutron activation. (author)

  16. Supply of enriched uranium for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, H. [NUKEM GmbH, Alzenau (Germany)

    1997-08-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

  17. Facility for a Low Power Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chalker, R. G.

    1949-09-14

    Preliminary investigation indicates that a reactor facility with ample research provisions for use by University or other interested groups, featuring safety in design, can be economically constructed in the Los Angeles area. The complete installation, including an underground gas-tight reactor building, with associated storage and experiment assembly building, administration offices, two general laboratory buildings, hot latoratory and lodge, can be constructed for approxinately $1,500,000. This does not include the cost of the reactor itself or of its auxiliary equipment,

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-10-29

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

  19. Strengthening IAEA Safeguards for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reid, Bruce D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anzelon, George A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Budlong-Sylvester, Kory [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-01

    During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half a dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors could pose a safeguards challenge. To strengthen the effectiveness of safeguards at the State level, this paper advocates that the IAEA consider ways to focus additional attention and broaden its safeguards toolbox for research reactors. This increase in focus on the research reactors could begin with the recognition that the research reactor (of any size) could be a common path element on a large number of technically plausible pathways that must be considered when performing acquisition pathway analysis (APA) for developing a State Level Approach (SLA) and Annual Implementation Plan (AIP). To

  20. Research on plasma core reactors

    Science.gov (United States)

    Jarvis, G. A.; Barton, D. M.; Helmick, H. H.; Bernard, W.; White, R. H.

    1976-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diameter by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17 cm thick by 89 cm diameter beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000 cu cm aluminum canister in the central region was fueled with UF6 gas and fission density distributions determined. These results are to be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

  1. Decommissioning of the Salaspils Research Reactor

    Directory of Open Access Journals (Sweden)

    Abramenkovs Andris

    2011-01-01

    Full Text Available In May 1995, the Latvian government decided to shut down the Salaspils Research Reactor and to dispense with nuclear energy in the future. The reactor has been out of operation since July 1998. A conceptual study on the decommissioning of the Salaspils Research Reactor was drawn up by Noell-KRC-Energie- und Umwelttechnik GmbH in 1998-1999. On October 26th, 1999, the Latvian government decided to start the direct dismantling to “green-field” in 2001. The upgrading of the decommissioning and dismantling plan was carried out from 2003-2004, resulting in a change of the primary goal of decommissioning. Collecting and conditioning of “historical” radioactive wastes from different storages outside and inside the reactor hall became the primary goal. All radioactive materials (more than 96 tons were conditioned for disposal in concrete containers at the radioactive wastes depository “Radons” at the Baldone site. Protective and radiation measurement equipment of the personnel was upgraded significantly. All non-radioactive equipment and materials outside the reactor buildings were released for clearance and dismantled for reuse or conventional disposal. Contaminated materials from the reactor hall were collected and removed for clearance measurements on a weekly basis.

  2. The current status of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tri Wulan Tjiptono; Syarip

    1998-10-01

    The Kartini reactor reached the first criticality on January 25, 1979. In the first three years, the reactor power is limited up to 50 kW thermal power and on July 1, 1982 has been increased to 100 kW. It has been used as experiments facility by researcher of Atomic Energy National Agency and students of the Universities. Three beam tubes used as experiments facilities, the first, is used as a neutron source for H{sub 2}O-Natural Uranium Subcritical Assembly, the second, is developed for neutron radiography facility and the third, is used for gamma radiography facility. The other facilities are rotary rack and two pneumatic transfer systems, one for delayed neutron counting system and the other for the new Neutron Activation Analysis (NAA) facility. The rotary rack used for isotope production for NAA purpose (for long time irradiation), the delayed neutron counting system used for analysis the Uranium contents of the ores and the new NAA is provided for short live elements analysis. In the last three years the Reactor Division has a joint use program with the Nuclear Component and Engineering Center in research reactor instrumentation and control development. (author)

  3. Radionuclide release from research reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Curtius, H., E-mail: h.curtius@fz-juelich.de [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany); Kaiser, G.; Mueller, E.; Bosbach, D. [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany)

    2011-09-01

    Numerous investigations with respect to LWR fuel under non oxidizing repository relevant conditions were performed. The results obtained indicate slow corrosion rates for the UO{sub 2} fuel matrix. Special fuel-types (mostly dispersed fuels, high enriched in {sup 235}U, cladded with aluminium) are used in German research reactors, whereas in German nuclear power plants, UO{sub 2}-fuel (LWR fuel, enrichment in {sup 235}U up to 5%, zircaloy as cladding) is used. Irradiated research reactor fuels contribute less than 1% to the total waste volume. In Germany, the state is responsible for fuel operation and for fuel back-end options. The institute for energy research (IEF-6) at the Research Center Juelich performs investigation with irradiated research reactor spent fuels under repository relevant conditions. In the study, the corrosion of research reactor spent fuel has been investigated in MgCl{sub 2}-rich salt brine and the radionuclide release fractions have been determined. Leaching experiments in brine with two different research reactor fuel-types were performed in a hot cell facility in order to determine the corrosion behaviour and the radionuclide release fractions. The corrosion of two dispersed research reactor fuel-types (UAl{sub x}-Al and U{sub 3}Si{sub 2}-Al) was studied in 400 mL MgCl{sub 2}-rich salt brine in the presence of Fe{sup 2+} under static and initially anoxic conditions. Within these experimental parameters, both fuel types corroded in the experimental time period of 3.5 years completely, and secondary alteration phases were formed. After complete corrosion of the used research reactor fuel samples, the inventories of Cs and Sr were quantitatively detected in solution. Solution concentrations of Am and Eu were lower than the solubility of Am(OH){sub 3}(s) and Eu(OH){sub 3}(s) solid phases respectively, and may be controlled by sorption processes. Pu concentrations may be controlled by Pu(IV) polymer species, but the presence of Pu(V) and Pu

  4. 78 FR 58575 - Review of Experiments for Research Reactors

    Science.gov (United States)

    2013-09-24

    ... COMMISSION Review of Experiments for Research Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Guide (RG) 2.4, ``Review of Experiments for Research Reactors.'' The guide is being withdrawn because... Experiments for Research Reactors,'' (ADAMS Accession No. ML003740131) because its guidance no longer provides...

  5. The AFR. An approved network of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hampel, Gabriele [Mainz Univ. (Germany). Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren (AFR)

    2012-10-15

    AFR (Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren) is the German acronym for 'Association for Research Reactor Operation and Safety Issues' which was founded in 1959. Reactor managers of European research reactors mainly from the German linguistic area meet regularly for their mutual benefit to exchange experience and knowledge in all areas of operating, managing and utilization of research reactors. In the last 2 years joint meetings were held together with the French association of research reactors CER (Club d'Exploitants des Reacteurs). In this contribution the AFR, its members, work and aims as well as the French partner CER are presented. (orig.)

  6. 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.

  7. Dynamic feedback characteristics of Ghana Research Reactor-1 ...

    African Journals Online (AJOL)

    Dynamic experiments were performed to investigate the effects of insertions of step and ramp reactivities on Ghana Research Reactor-1. These safety performance tests of the reactor show that the reactor is inherently safe. The peak powers were found to be low and could not lead to damage of fuel meat and cladding.

  8. Meteodiffusive Characterization of Algiers' Nuclear Research Reactor

    Directory of Open Access Journals (Sweden)

    Mourad Messaci

    2007-01-01

    Full Text Available In the framework of the environmental impact studies of the nuclear research reactor of Algiers, we will present the work related to the atmospheric dispersion of releases due to the installation in normal operation, which dealt with the assessment of spatial distribution of yearly average values of atmospheric dilution factor. The aim of this work is a characterization of the site in terms of diffusivity, which is basic for the radiological impact evaluation of the reactor. The meteorological statistics result from the National Office of Meteorology and concern 15 years of hourly records. According to the nature and features of these data, a Gaussian-type model with wind direction sectors was used. Values of wind speed at release height were estimated from measurement values at 10 m from ground. For the assessment of vertical dispersion coefficient, we used Briggs' formulas related to a sampling time of one hour. Areas of maximum impact were delimited and points of highest concentration within these zones were identified.

  9. Research on the reactor physics and reactor safety of VVER reactors. AER Symposium 2016

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, S.

    2017-09-15

    The selected paperscan be attributed to the following main subjects: Reactor start-up tests and use of corresponding data for code validation, code development and application, approaches for safety analyses, closure of nuclear fuel cycle, prospective reactor concepts.

  10. The current status of nuclear research reactor in Thailand

    Energy Technology Data Exchange (ETDEWEB)

    Sittichai, C.; Kanyukt, R.; Pongpat, P. [Office of Atomic Energy for Peace, Bangkok (Thailand)

    1998-10-01

    Since 1962, the Thai Research Reactor has been serving for various kinds of activities i.e. the production of radioisotopes for medical uses and research and development on nuclear science and technology, for more than three decades. The existing reactor site should be abandoned and relocated to the new suitable site, according to Thai cabinet`s resolution on the 27 December 1989. The decommissioning project for the present reactor as well as the establishment of new nuclear research center were planned. This paper discussed the OAEP concept for the decommissioning programme and the general description of the new research reactor and some related information were also reported. (author)

  11. The research reactors their contribution to the reactors physics; Les reacteurs de recherche leur apport sur la physique des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Zaetta, A. [CEA/Cadarache, Direction des Reacteurs Nucleaires, DRN, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint Paul lez Durance (France); Mathoniere, G. [CEA/Saclay, DEN, 91 - Gif sur Yvette (France)] [and others

    2000-07-01

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  12. Diversion assumptions for high-powered research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Binford, F.T.

    1984-01-01

    This study deals with diversion assumptions for high-powered research reactors -- specifically, MTR fuel; pool- or tank-type research reactors with light-water moderator; and water, beryllium, or graphite reflectors, and which have a power level of 25 MW(t) or more. The objective is to provide assistance to the IAEA in documentation of criteria and inspection observables related to undeclared plutonium production in the reactors described above, including: criteria for undeclared plutonium production, necessary design information for implementation of these criteria, verification guidelines including neutron physics and heat transfer, and safeguards measures to facilitate the detection of undeclared plutonium production at large research reactors.

  13. Initial decommissioning planning for the Budapest research reactor

    Directory of Open Access Journals (Sweden)

    Toth Gabor

    2011-01-01

    Full Text Available The Budapest Research Reactor is the first nuclear research facility in Hungary. The reactor is to remain in operation for at least another 13 years. At the same time, the development of a decommissioning plan is a mandatory requirement under national legislation. The present paper describes the current status of decommissioning planning which is aimed at a timely preparation for the forthcoming decommissioning of the reactor.

  14. Photoneutron effects on pulse reactor kinetics for the Annular Core Research Reactor (ACRR).

    Energy Technology Data Exchange (ETDEWEB)

    Parma, Edward J., Jr.

    2009-06-01

    The Annular Core Research Reactor (ACRR) is a swimming-pool type pulsed reactor that maintains an epithermal neutron flux and a nine-inch diameter central dry cavity. One of its uses is neutron and gamma-ray irradiation damage studies on electronic components under transient reactor power conditions. In analyzing the experimental results, careful attention must be paid to the kinetics associated with the reactor to ensure that the transient behavior of the electronic device is understood. Since the ACRR fuel maintains a substantial amount of beryllium, copious quantities of photoneutrons are produced that can significantly alter the expected behavior of the reactor power, especially following a reactor pulse. In order to understand these photoneutron effects on the reactor kinetics, the KIFLE transient reactor-analysis code was modified to include the photoneutron groups associated with the beryllium. The time-dependent behavior of the reactor power was analyzed for small and large pulses, assuming several initial conditions including following several pulses during the day, and following a long steady-state power run. The results indicate that, for these types of initial conditions, the photoneutron contribution to the reactor pulse energy can have a few to tens of percent effect.

  15. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective ...

  16. Proceedings of the sixth Asian symposium on research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-08-01

    The symposium consisted of 16 sessions with 58 submitted papers. Major fields were: (1) status and future plan of research and testing reactors, (2) operating experiences, (3) design and modification of the facility, and reactor fuels, (4) irradiation studies, (5) irradiation facilities, (6) reactor characteristics and instrumentation, and (7) neutron beam utilization. Panel discussion on the 'New Trends on Application of Research and Test Reactors' was also held at the last of the symposium. About 180 people participated from China, Korea, Indonesia, Thailand, Bangladesh, Vietnam, Chinese Taipei, Belgium, France, USA, Japan and IAEA. The 58 of the presented papers are indexed individually. (J.P.N.)

  17. Antineutrino and gamma emission from the OSIRIS research reactor

    Science.gov (United States)

    Giot, Lydie; Fallot, Muriel

    2017-09-01

    For the first time, the summation method has been coupled with a complete reactor model, in order to predict the antineutrino emission of a research reactor. This work, discussed in the first part of this paper, allows us to predict the low energy part of the antineutrino spectrum, evidencing the important contribution of actinides to the antineutrino emission. Experimental conditions at short distance from research reactors are challenging, because the reactor itself produces huge gamma background that induce accidental and correlated backgrounds in an antineutrino target. The understanding of this background is of utmost importance and triggered the second part of the work presented here.

  18. Related activities on management of ageing of Dalat Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pham Van Lam [Reactor Dept., Nuclear Research Institute, Dalat (Viet Nam)

    1998-10-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the previous 250 kW TRIGA-MARK II reactor. The reactor core, the control and instrumentation system, the primary and secondary cooling systems as well as other associated systems were newly designed and installed. The renovated reactor reached its initial criticality in November 1983 and attained its nominal power of 500 kW in February 1984. Since then DNRR has been operated safely. Retained structures of the former reactor such as the reactor aluminum tank, the graphite reflector, the thermal column, the horizontal beam tubes and the radiation concrete shielding are 35 years old. During the recent years, in-service inspection has been carried out, the reactor control and instrumentation system were renovated due to ageing and obsolescence of its components, reactor general inspection and refurbishment were performed. Efforts are being made to cope with ageing of old reactor components to maintain safe operation of the DNRR. (author)

  19. Sodium fast reactor safety and licensing research plan. Volume II.

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  20. 75 FR 70042 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

    Science.gov (United States)

    2010-11-16

    ... All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order... above, shall be submitted to the NRC to the attention of the Director, Office of Nuclear Reactor... properly marked and handled in accordance with 10 CFR 73.21. The Director, Office of Nuclear Reactor...

  1. 75 FR 79423 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

    Science.gov (United States)

    2010-12-20

    ... All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order... Director, Office of Nuclear Reactor Regulation under 10 CFR 50.4. In addition, licensee submittals that... Director, Office of Nuclear Reactor Regulation, may, in writing, relax or rescind any of the above...

  2. A Potential NASA Research Reactor to Support NTR Development

    Science.gov (United States)

    Eades, Michael; Gerrish, Harold; Hardin, Leroy

    2013-01-01

    In support of efforts for research into the design and development of a man rated Nuclear Thermal Rocket (NTR) engine, the National Aeronautics and Space Administration (NASA), Marshall Space Flight Center (MSFC), is evaluating the potential for building a Nuclear Regulatory Commission (NRC) licensed research reactor. The proposed reactor would be licensed by NASA and operated jointly by NASA and university partners. The purpose of this reactor would be to perform further research into the technologies and systems needed for a successful NTR project and promote nuclear training and education.

  3. JPL in-house fluidized-bed reactor research

    Science.gov (United States)

    Rohatgi, N. K.

    1984-01-01

    Fluidized bed reactor research techniques for fabrication of quartz linears was reviewed. Silane pyrolysis was employed in this fabrication study. Metallic contaminant levels in the silicon particles were below levels detectable by emission spectroscopy.

  4. Renewing Liquid Fueled Molten Salt Reactor Research and Development

    Science.gov (United States)

    Towell, Rusty; NEXT Lab Team

    2016-09-01

    Globally there is a desperate need for affordable, safe, and clean energy on demand. More than anything else, this would raise the living conditions of those in poverty around the world. An advanced reactor that utilizes liquid fuel and molten salts is capable of meeting these needs. Although, this technology was demonstrated in the Molten Salt Reactor Experiment (MSRE) at ORNL in the 60's, little progress has been made since the program was cancelled over 40 years ago. A new research effort has been initiated to advance the technical readiness level of key reactor components. This presentation will explain the motivation and initial steps for this new research initiative.

  5. A Design of Alarm System in a Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaekwan; Jang, Gwisook; Seo, Sangmun; Suh, Yongsuk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The digital alarm system has become an indispensable design to process a large amount of alarms of power plants. Korean research reactor operated for decades maintains a hybrid alarm system with both an analog annunciator and a digital alarm display. In this design, several alarms are indicated on an analog panel and digital display, respectively, and it requires more attention and effort of the operators. As proven in power plants, a centralized alarm system design is necessary for a new research reactor. However, the number of alarms and operators in a research reactor is significantly lesser than power plants. Thus, simplification should be considered as an important factor for the operation efficiency. This paper introduces a simplified alarm system. As advances in information technology, fully digitalized alarm systems have been applied to power plants. In a new research reactor, it will be more useful than an analog or hybrid configuration installed in research reactors decades ago. However, the simplification feature should be considered as an important factor because the number of alarms and number of operators in a research reactor is significantly lesser than in power plants.

  6. Development of Digital MMIS for Research Reactors: Graded Approaches

    Energy Technology Data Exchange (ETDEWEB)

    Khalil ur, Rahman; Shin, Jin Soo; Heo, Gyun Young [Kyunghee University, Yongin (Korea, Republic of); Son, Han Seong [Joongbu University, Geumsan (Korea, Republic of); Kim, Young Ki; Park, Jae Kwan; Seo, Sang Mun; Kim, Yong Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Though research reactors are small in size yet they are important in terms of industrial applications and R and D, educational purposes. Keeping the eye on its importance, Korean government has intention to upgrade and extend this industry. Presently, Korea is operating only HANARO at Korea Atomic Energy Research Institute (KAERI) and AGN-201K at Kyung Hee University (KHU), which are not sufficient to meet the current requirements of research and education. In addition, we need self-sufficiency in design and selfreliance in design and operation, as we are installing research reactors in domestic as well as foreign territories for instance Jordan. Based on these demands, KAERI and universities initiated a 5 year research project since December 2011 collaboratly, for the deep study of reactor core, thermal hydraulics, materials and instrumentation and control (I and C). This particular study is being carried out to develop highly reliable advanced digital I and C systems using a grading approach. It is worth mentioning that next generation research reactor should be equipped with advance state of the art digital I and C for safe and reliable operation and impermeable cyber security system that is needed to be devised. Moreover, human error is one of important area which should be linked with I and C in terms of Man Machine Interface System (MMIS) and development of I and C should cover human factor engineering. Presently, the digital I and C and MMIS are well developed for commercial power stations whereas such level of development does not exist for research reactors in Korea. Since the functional and safety requirements of research reactors are not so strict as commercial power plants, the design of digital I and C systems for research reactors seems to be graded based on the stringency of regulatory requirements. This paper was motivated for the introduction of those missions, so it is going to describe the general overview of digital I and C systems, the graded

  7. Activities for extending the lifetime of MINT research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, Adnan; Kassim, Mohammad Suhaimi [Malaysian Inst. for Nuclear Technology Research (MINT), Bangi, Kajang (Malaysia)

    1998-10-01

    MINT TRIGA Reactor is a 1-MW swimming pool nuclear reactor commissioned in June 1982. Since then, it has been used for research, isotope production, neutron activation, neutron radiography and manpower training. The total operating time till the end on September 1997 is 16968 hours with cumulative total energy release of 11188 MW-hours. After more than fifteen years of successful operation, some deterioration in components and associated systems has been observed. This paper describes some of the activities carried out to increase the lifetime and to reduce the shutdown time of the reactor. (author)

  8. Development of a research nuclear reactor simulator using LABVIEW®

    Energy Technology Data Exchange (ETDEWEB)

    Lage, Aldo Marcio Fonseca; Mesquita, Amir Zacarias; Pinto, Antonio Juscelino; Souza, Luiz Claudio Andrade [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. The most important variable in the nuclear reactors control is the power released by fission of the fuel in the core which is directly proportional to neutron flux. It was developed a digital system to simulate the neutron evolution flux and monitoring their interaction on the other operational parameters. The control objective is to bring the reactor power from its source level (mW) to a few W. It is intended for education of basic reactor neutronic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron and control by rods. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Center - CDTN (Belo Horizonte/Brazil) was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. They are cooled by light water under natural convection and are characterized by being inherently safety. The simulation system was developed using the LabVIEW® (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's). The main purpose of the system is to provide to analyze the behavior, and the tendency of some processes that occur in the reactor using a user-friendly operator interface. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility.(author)

  9. Reactor Safety Research Programs Quarterly Report April- June 1981

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1981-09-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL} from April1 through June 30, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory {INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  10. Operation of the FRG-research reactors at Geesthacht

    Energy Technology Data Exchange (ETDEWEB)

    Reymann, A.; Krull, W. [GKSS-Forschungszentrum Geesthacht GmbH, Geesthacht-Tesperhude (Germany)

    1997-07-01

    Two research reactors have been operated very successfully by the GKSS research centre over decades in a connected pool system. FRG-1: 5 MW, criticality October 1958, FRG-2: 15 MW, criticality March 1963 and decommissioned February 1995. The FRG-2 was scheduled to stop operation in 1991 for lack of scientific and technical interest for future use. The reactor has been used as Germany's largest material testing reactor for power reactor fuel and power reactor materials development and safety tests. The FRG-2 has also played an important role in the conversion activities at the GKSS research centre. The FRG-1 is being used with high availability for beam tube experiments for fundamental and applied research in biology, membrane development, materials research, neutron radiography, neutron activation analyses etc. To enable the sufficient and efficient use of long wave length neutrons a cold neutron source has been installed in one of the beam tubes. The GKSS research centre, the advisory board, the scientific community and other clients are demanding the ongoing operation of the FRG-1 for at least till the year 2010. GKSS has taken many actions to ensure the save operation with high utilization and availability for the next 15 years. (author)

  11. Sustainability management for operating organizations of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kibrit, Eduardo; Aquino, Afonso Rodrigues de, E-mail: ekibrit@ipen.br, E-mail: araquino@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    Sustainable development is development that meets the needs of the present without compromising the ability of future generations to meet their own needs. In a country like Brazil, where nuclear activity is geared towards peaceful purposes, any operating organization of research reactor should emphasize its commitment to social, environmental, economic and institutional aspects. Social aspects include research and development, production and supply of radiopharmaceuticals, radiation safety and special training for the nuclear sector. Environmental aspects include control of the surroundings and knowledge directed towards environment preservation. Economic aspects include import substitution and diversification of production. Institutional aspects include technology, innovation and knowledge. These aspects, if considered in the management system of an operating organization of research reactor, will help with its long-term maintenance and success in an increasingly competitive market scenario. About this, we propose a sustainability management system approach for operating organizations of research reactors. A bibliographical review on the theme is made. A methodology for identifying indicators for measuring sustainability in nuclear research reactors processes is also described. Finally, we propose a methodology for sustainability perception assessment to be applied at operating organizations of research reactors. (author)

  12. The rehabilitation/upgrading of Philippine Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Renato, T. Banaga [Philippines Nuclear Research Inst., Quezon (Philippines)

    1998-10-01

    The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E{sub 1}-U-Z{sub 1}-H{sub 1.6} TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

  13. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

  14. Radiation protection personnel training in Research Reactors; Capacitacion en proteccion radiologica para reactores de investigacion

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, Carlos Dario; Lorenzo, Nestor Pedro de [Comision Nacional de Energia Atomica, Rio Negro (Argentina). Centro Atomico Bariloche. Instituto Balseiro

    1996-07-01

    The RA-6 research reactor is considering the main laboratory in the training of different groups related with radiological protection. The methodology applied to several courses over 15 years of experience is shown in this work. The reactor is also involved in the construction, design, start-up and sell of different installation outside Argentina for this reason several theoretical and practical courses had been developed. The acquired experience obtained is shown in this paper and the main purpose is to show the requirements to be taken into account for every group (subjects, goals, on-job training, etc) (author)

  15. Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Baek, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, L-Y [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-01-30

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

  16. Reactor Safety Research Programs Quarterly Report October - December 1980

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S K

    1981-04-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from October 1 through December 31, 1980, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NOE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  17. Reactor Safety Research Programs Quarterly Report July- September 1980

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1980-12-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  18. Reactor Safety Research Programs Quarterly Report April -June 1980

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1980-11-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  19. Reactor Safety Research: Semiannual report, July-December 1986

    Energy Technology Data Exchange (ETDEWEB)

    1987-11-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the tehcnology base supporting licensing decisions.

  20. Study on the Export Strategies for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oh, S. K.; Lee, Y. J.; Ham, T. K.; Hong, S. T.; Kim, J. H. [Ajou University, Suwon (Korea, Republic of)

    2008-12-15

    Key strategic considerations taken into account should be based on understanding in the forecasts of demand and supply balance as well as the missions of research reactor for customers. For timely arrival at the competition, it may be advantageous to categorize the potential customers into 3 groups, the developed, the developing and the underdeveloped countries in respect of nuclear technology, and to be ready for the group-wise reference designs of the key reactor systems. Customizing the design to specific owner's requirements can advance from one of these reference designs when competition starts. To mobilize this approach effectively, it is useful to establish an integral project and technology management system earlier. This system will function as an important success factor for international research reactor business, because it makes easy to accommodate customer requirements and to achieve the design-to-cost.

  1. Dismantling of the DIORIT research reactor - Conditioning of activated graphite.

    Science.gov (United States)

    Sierra Perler, Isabel Cecilia; Beer, Hans-Frieder; Müth, Joachim; Kramer, Andreas

    2017-08-16

    The research reactor DIORIT at the Paul Scherrer Institute was a natural uranium reactor moderated by D2O. It was put in operation in 1960 and finally shut down in August 1977. The dismantling project started in 1982 and could be successfully finished on September 11th, 2012. About 40 tons of activated reactor graphite had to be conditioned during the dismantling of this research reactor. The problem of conditioning of activated reactor graphite had not been solved so far worldwide. Therefore a conditioning method considering radiation protection and economic aspects had to be developed. As a result, the graphite was crushed to a particle size smaller than 5 mm and added as sand substitute to a specially developed grout. The produced graphite concrete was used as a matrix for embedding dismantling waste in containers. By conditioning the graphite conventionally, about 58.5 m3 (13 containers) of waste volume would have been generated. The new PSI invention resulted in no additional waste caused by graphite. Consequently, the resulting waste volume, as well as the costs, were substantially reduced. Copyright © 2017. Published by Elsevier Ltd.

  2. 77 FR 26321 - Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112

    Science.gov (United States)

    2012-05-03

    ... COMMISSION Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112 AGENCY..., Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Rockville, MD 20852. Telephone..., Research and Test Reactors Licensing Branch, Division of Policy and Rulemaking, Office of Nuclear Reactor...

  3. Best Safety Practices for the Operation of Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Boeck, H.; Villa, M. [Atominstitute of the Austrian Universities, 1020 Vienna (Austria)

    2002-07-01

    A survey on administrative, organisational and technical aspects for the safe and efficient operation of a 250 kW TRIGA Mark II research reactor is given. The replacement of the I and C system is discussed, maintenance procedures are presented and the fuel management is described. (author)

  4. Neutron spectrometric methods for core inventory verification in research reactors

    CERN Document Server

    Ellinger, A; Hansen, W; Knorr, J; Schneider, R

    2002-01-01

    In consequence of the Non-Proliferation Treaty safeguards, inspections are periodically made in nuclear facilities by the IAEA and the EURATOM Safeguards Directorate. The inspection methods are permanently improved. Therefore, the Core Inventory Verification method is being developed as an indirect method for the verification of the core inventory and to check the declared operation of research reactors.

  5. Epiboron Neutron Activation Analysis with Nigeria Research Reactor

    African Journals Online (AJOL)

    Epiboron neutron activation analysis is optimized using Nigeria Research Reactor-1. Data are given for 6 elements using boron as shielding. Boron shield are of particular practical value for rapid instrumental analysis. Advantage factors for the following elements: I, Br, Cl, K, Mn and Na under boron shield are given.

  6. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  7. Fault detection system for Argentine Research Reactor instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Polenta, H.P. (Argentine Navy, Comodoro Py 2055 Office 11-93, 1104 - Buenos Aires (Argentina)); Bernard, J.A. (Nuclear Reactor Laboratory, Massachusetts Institute of Technology, 138 Albany Street, Cambridge, Massachusetts 02139 (United States)); Ray, A. (205 Mechanical Engineering Department, Pennsylvania State University, University Park, Pennsylvania 16802 (United States))

    1993-01-20

    The design and implementation of a redundancy management scheme for the on-line detection and isolation of faulty sensors is presented. Such a device is potentially useful in reactor-powered spacecraft for enhancing the processing capabilities of the main computer. The fault detection device can be used as an integral part of intelligent instrumentation systems. The device has been built using an 8-bit microcontroller and commercially available electronic hardware. The software is completely portable. The operation of this device has been successfully demonstrated for real-time validation of sensor data on Argentina's RA-1 Research Reactor.

  8. Fuel shuffling optimization for the Delft research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Geemert, R. van; Hoogenboom, J.E.; Gibcus, H.P.M. [Delft Univ. of Technology, Interfaculty Reactor Inst., Delft (Netherlands); Quist, A.J. [Delft Univ., Fac. of Applied Mathematics and Informatics, Delft (Netherlands)

    1997-07-01

    A fuel shuffling optimization procedure is proposed for the Hoger Onderwijs Reactor (HOR) in Delft, the Netherlands, a 2 MWth swimming-pool type research reactor. In order to cope with the fluctuatory behaviour of objective functions in loading pattern optimization, the proposed cyclic permutation optimization procedure features a gradual transition from global to local search behaviour via the introduction of stochastic tests for the number of fuel assemblies involved in a cyclic permutation. The possible objectives and the safety and operation constraints, as well as the optimization procedure, are discussed, followed by some optimization results for the HOR. (author)

  9. Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

    Energy Technology Data Exchange (ETDEWEB)

    Heeger, Karsten M. [Yale Univ., New Haven, CT (United States)

    2014-09-13

    This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zero$\\theta_{13}$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.

  10. Neutron beams implemented at nuclear research reactors for BNCT

    Science.gov (United States)

    Bavarnegin, E.; Kasesaz, Y.; Wagner, F. M.

    2017-05-01

    This paper presents a survey of neutron beams which were or are in use at 56 Nuclear Research Reactors (NRRs) in order to be used for BNCT, either for treatment or research purposes in aspects of various combinations of materials that were used in their Beam Shaping Assembly (BSA) design, use of fission converters and optimized beam parameters. All our knowledge about BNCT is indebted to researches that have been done in NRRs. The results of about 60 years research in BNCT and also the successes of this method in medical treatment of tumors show that, for the development of BNCT as a routine cancer therapy method, hospital-based neutron sources are needed. Achieving a physical data collection on BNCT neutron beams based on NRRs will be helpful for beam designers in developing a non-reactor based neutron beam.

  11. Needs and Requirements for Future Research Reactors (ORNL Perspectives)

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bryan, Chris [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gehin, Jess C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-02-10

    The High Flux Isotope Reactor (HFIR) is a vital national and international resource for neutron science research, production of radioisotopes, and materials irradiation. While HFIR is expected to continue operation for the foreseeable future, interest is growing in understanding future research reactors features, needs, and requirements. To clarify, discuss, and compile these needs from the perspective of Oak Ridge National Laboratory (ORNL) research and development (R&D) missions, a workshop, titled “Needs and Requirements for Future Research Reactors”, was held at ORNL on May 12, 2015. The workshop engaged ORNL staff that is directly involved in research using HFIR to collect valuable input on the reactor’s current and future missions. The workshop provided an interactive forum for a fruitful exchange of opinions, and included a mix of short presentations and open discussions. ORNL staff members made 15 technical presentations based on their experience and areas of expertise, and discussed those capabilities of the HFIR and future research reactors that are essential for their current and future R&D needs. The workshop was attended by approximately 60 participants from three ORNL directorates. The agenda is included in Appendix A. This document summarizes the feedback provided by workshop contributors and participants. It also includes information and insights addressing key points that originated from the dialogue started at the workshop. A general overview is provided on the design features and capabilities of high performance research reactors currently in use or under construction worldwide. Recent and ongoing design efforts in the US and internationally are briefly summarized, followed by conclusions and recommendations.

  12. Comparative Study on Cyber Securities between Power Reactor and Research Reactor with Bayesian Update

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Jinsoo; Heo, Gyunyoung [Kyung Hee University, Yongin (Korea, Republic of); Son, Han Seong [Joongbu Univiersity, Geumsan (Korea, Republic of)

    2016-10-15

    The Stuxnet has shown that nuclear facilities are no more safe from cyber-attack. Due to practical experiences and concerns on increasing of digital system application, cyber security has become the important issue in nuclear industry. Korea Institute of Nuclear Nonproliferation and control (KINAC) published a regulatory standard (KINAC/RS-015) to establish cyber security framework for nuclear facilities. However, it is difficult to research about cyber security. It is hard to quantify cyber-attack which has malicious activity which is different from existing design basis accidents (DBAs). We previously proposed a methodology on development of a cyber security risk model with BBN. However, the methodology had a limitation in which the input data as prior information was solely on expert opinions. In this study, we propose a cyber security risk model for instrumentation and control (I and C) system of nuclear facilities with some equation for quantification by using Bayesian Belief Network (BBN) in order to overcome the limitation of previous research. The proposed model has been used for comparative study on cyber securities between large-sized nuclear power plants (NPPs) and small-sized Research Reactors (RR). In this study, we proposed the cyber security risk evaluation model with BBN. It includes I and C architecture, which is a target system of cyber-attack, malicious activity, which causes cyber-attack from attacker, and mitigation measure, which mitigates the cyber-attack risk. Likelihood and consequence as prior information are evaluated by considering characteristics of I and C architecture and malicious activity. The BBN model provides posterior information with Bayesian update by adding any of assumed cyber-attack scenarios as evidence. Cyber security risk for nuclear facilities is analyzed by comparing between prior information and posterior information of each node. In this study, we conducted comparative study on cyber securities between power reactor

  13. 75 FR 57080 - In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order...

    Science.gov (United States)

    2010-09-17

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order... which authorizes the possession, use, and operation of the Aerotest Radiography and Research Reactor...

  14. Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-04-20

    The United States Department of Energy (DOE) commissioned the development of technology roadmaps for advanced (non-light water reactor) reactor concepts to help focus research and development funding over the next five years. The roadmaps show the research and development needed to support demonstration of an advanced (non-LWR) concept by the early 2030s, consistent with DOE’s Vision and Strategy for the Development and Deployment of Advanced Reactors. The intent is only to convey the technical steps that would be required to achieve such a goal; the means by which DOE will determine whether to invest in specific tasks will be treated separately. The starting point for the roadmaps is the Technical Readiness Assessment performed as part of an Advanced Test and Demonstration Reactor study released in 2016. The roadmaps were developed based upon a review of technical reports and vendor literature summarizing the technical maturity of each concept and the outstanding research and development needs. Critical path tasks for specific systems were highlighted on the basis of time and resources needed to complete the tasks and the importance of the system to the performance of the reactor concept. The roadmaps are generic, i.e. not specific to a particular vendor’s design but vendor design information may have been used as representative of the concept family. In the event that both near-term and more advanced versions of a concept are being developed, either a single roadmap with multiple branches or separate roadmaps for each version were developed. In each case, roadmaps point to a demonstration reactor (engineering or commercial) and show the activities that must be completed in parallel to support that demonstration in the 2030-2035 window. This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The SFR technology is mature enough for commercial demonstration by the early 2030s

  15. Reactor-produced radionuclides at the University of Missouri Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ketring, A.R.; Evans-Blumer, M.S.; Ehrhardt, G.J. [University of Missouri Research Reactor, Colombia (United States). Departments of Radiology, Chemistry and Nuclear Engineering

    1997-10-01

    Nuclear medicine has primarily been a diagnostic science for many years, but today is facing considerable challenges from other modalities in this area. However, these competing techniques (magnetic resonance imaging, ultrasound, and computer-assisted tomography) in general are not therapeutic. Although early nuclear medicine therapy was of limited efficacy, in recent years a revolution in radiotherapy has been developing base don more sophisticated targeting methods, including radioactive intra-arterial microspheres, chemically-guided bone agents, labelled monoclonal antibodies, and isotopically-tagged polypeptide receptor-binding agents. Although primarily used for malignancies, therapeutic nuclear medicine is also applicable to the treatment of rheumatoid arthritis and possibly coronary artery re closure following angioplasty. The isotopes of choice for these applications are reactor-produced beta emitters such as Sm-153, Re-186, Re-188, Ho-166, Lu-177, and Rh-105. Although alpha emitters possess greater cell toxicity due to their high LET, the greater range of beta emitters and the typically inhomogeneous deposition of radiotherapy agents in lesions leads to greater beta `crossfire` and better overall results. The University of Missouri Research Reactor (MURR) has been in the forefront of research into means of preparing, handling and supplying these high-specific-activity isotopes in quantities appropriate not only for research, but also for patient trials in the US and around the world. Researchers at MURR in collaboration with others at the University of Missouri (MU) developed Sm-153 Quadramet{sup TM}, a drug recently approved in the US for palliation of bone tumor pain. In conjunction with researchers at the University of Missouri-Rolla, MURR also developed Y-90 TheraSphere{sup TM}, an agent for the treatment of liver cancer now approved in Canada. Considerable effort has been expended to develop techniques for irradiation, handling, and shipping isotopes

  16. IAEA designated international centre based on research reactors (ICERR)

    Energy Technology Data Exchange (ETDEWEB)

    Di Tigliole, Andrea Borio; Bradley, Edward; Khoroshev, Mikhail; Marshall, Frances; Morris, Charles; Tozser, Sandor [International Atomic Energy Agency, Vienna (Austria). Dept. of Nuclear Energy

    2016-04-15

    International activities in the back end of the research reactor (RR) fuel cycle have so far been dominated by the programmes of acceptance of highly-enriched uranium (HEU) spent nuclear fuel (SNF) by the country where it was originally enriched. These programmes will soon have achieved their goals. However, the needs of the nuclear community dictate that the majority of the research reactors continues to operate using low enriched uranium (LEU) fuel in order to meet the varied mission objectives. As a result, inventories of LEU SNF will continue to be created and the back end solution of RR SNF remains a critical issue. In view of this fact, the IAEA drew up a report presenting available reprocessing and recycling services for RR SNF.

  17. Safety culture and quality management of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia); Hauptmanns, Ulrich [Department of Plant Design and Safety, Otto-Von-Guericke-University, Magdeburg (Germany)

    1999-10-01

    The evaluation for assessing the safety culture and quality of safety management of Kartini research reactor is presented. The method is based on the concept of management control of safety (audit) as well as by using the developed method i.e. the questionnaires concerning areas of relevance which have to be answered with value statements. There are seven statements or qualifiers in answering the questions. Since such statements are vague, they are represented by fuzzy numbers. The weaknesses can be identified from the different areas contemplated. The evaluation result show that the quality of safety management of Kartini research reactor is globally rated as 'Average'. The operator behavior in the implementation of 'safety culture' concept is found as a weakness, therefore this area should be improved. (author)

  18. 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, Division...

  19. Characterization of radioactive aerosols in Tehran research reactor containment

    Directory of Open Access Journals (Sweden)

    Moradi Gholamreza

    2015-01-01

    Full Text Available The objectives of this research were to determine the levels of radioactivity in the Tehran research reactor containment and to investigate the mass-size distribution, composition, and concentration of radionuclides during operation of the reactor. A cascade impactor sampler was used to determine the size-activity distributions of radioactive aerosols in each of the sampling stations. Levels of a and b activities were determined based on a counting method using a liquid scintillation counter and smear tests. The total average mass fractions of fine particles (particle diameter dp < 1 mm in all of the sampling stations were approximately 26.75 %, with the mean and standard deviation of 52.15 ± 19.75 mg/m3. The total average mass fractions of coarse particles were approximately 73.2%, with the mean and standard deviation of 71.34 ± 24.57 mg/m3. In addition to natural radionuclides, artificial radionuclides, such as 24Na, 91Sr, 131I, 133I, 103Ru, 82Br, and 140La, may be released into the reactor containment structure. Maximum activity was associated with accumulation-mode particles with diameters less than 400 nm. The results obtained from liquid scintillation counting suggested that the mean specific activity of alpha particles in fine and coarse-modes were 89.7 % and 10.26 %, respectively. The mean specific activity of beta particles in fine and coarse-modes were 81.15 % and 18.51 %, respectively. A large fraction of the radionuclides' mass concentration in the Tehran research reactor containment was associated with coarse-mode particles, in addition, a large fraction of the activity in the aerosol particles was associated with accumulation-mode particles.

  20. The TRIGA Reactor Facility at the Armed Forces Radiobiology Research Institute: A Simplified Technical Description.

    Science.gov (United States)

    1986-05-01

    AD-AiS68 238 THE TRIGA REACTOR FACILITY AT THE ARMED FORCESI! RADIOBIOLOGY RESEARCH INST..(U) ARMED FORCES RADIOBIOLOGY RESEARCH INST BETHESDA NO...medium-power exposure. The reactor is also used to train military personnel in reactor operations., The AFRRI TRIGA Mark-F reactor facility is within the...AFRRI complex on the grounds ’of the Naval Medical Command National Capital Region, in Bethesda, Maryland.> TRIGA is an acronym for Training, Research

  1. Feasibility of Thermoelectric Waste Heat Recovery from Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byunghee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    A thermoelectric generator has the most competitive method to regenerate the waste heat from research reactors, because it has no limitation on operating temperature. In addition, since the TEG is a solid energy conversion device converting heat to electricity directly without moving parts, the regenerating power system becomes simple and highly reliable. In this regard, a waste heat recovery using thermoelectric generator (TEG) from 15-MW pool type research reactor is suggested and the feasibility is demonstrated. The producible power from waste heat is estimated with respect to the reactor parameters, and an application of the regenerated power is suggested by performing a safety analysis with the power. The producible power from TEG is estimated with respect to the LMTD of the HX and the required heat exchange area is also calculated. By increasing LMTD from 2 K to 20K, the efficiency and the power increases greatly. Also an application of the power regeneration system is suggested by performing a safety analysis with the system, and comparing the results with reference case without the power regeneration.

  2. Developing strategic plans for effective utilization of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ridikas, Danas [International Atomic Energy Agency, Vienna (Austria). Dept. of Nuclear Sciences and Applications

    2015-12-15

    Strategic plans are indispensable documents for research reactors (RRs) to ensure their efficient, optimized and well managed utilization. A strategic plan provides a framework for increasing utilization, while helping to create a positive safety culture, a motivated staff, a clear understanding of real costs and a balanced budget. A strategic plan should be seen as an essential tool for a responsible manager of any RR, from the smallest critical facility to the largest reactor. Results and lessons learned are shown from the IAEA efforts to help the RR facilities developing strategic plans, provide review and advise services, organize national and regional stakeholder/user workshops, prepare further guidance and recommendations, document and publish guidance documents and other supporting materials.

  3. Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

    Energy Technology Data Exchange (ETDEWEB)

    Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics

    2017-05-15

    Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.

  4. Present status and future perspectives of research and test reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kaneko, Yoshihiko [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan); Kaieda, Keisuke [Department of Research Reactor, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-10-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

  5. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    Energy Technology Data Exchange (ETDEWEB)

    Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

    1999-08-01

    RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

  6. Contribution of CAD and PLM Research Reactors Design and Construction

    Energy Technology Data Exchange (ETDEWEB)

    Bonnetain, Xavier [AREVA TA, Paris (France)

    2013-07-01

    As all the reactors, the main stakes in the engineering of design and construction of the research reactors consist of the management and sharing of the technical data, the functional, physical and contractual interfaces data between the various contributors on the whole designs and construction cycle project. For 40 years, AREVA TA designs and builds reactors. Computer Aided Design (CAD) tools were introduced for 30 years into the engineering processes of AREVA TA, completed for 15 years by Product Lifecycle Management (PLM) tools. For 15 years AREVA TA pursues the integration since the feasibility of its newest Information Technologies (IT). In the first part, the paper presents IN the second part, the paper presents how the schematics and CAD tools support the engineering processes during the different phases of the project. CAD was used during the studies and now supports the management of the layout and design studies, including interfaces between suppliers, up to the constitution of the as built CAD mock-up. In the third part, the paper presents the relations between the various tools and the PLM solution implemented by AREVA TA to ensure the consistency between all tools and data for the benefit of the project.

  7. Joint KAERI/VAEC pre-possibility study on a new research reactor for Vietnam

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, B. C.; Chae, H. T.; Kim, H.; Lee, C. S.; Choi, C. O.; Jun, B. J. [KAERI, Taejon (Korea, Republic of); Vien, Luong Ba; Dien, Nguyen Nhi [Vietnam Atomic Energy Commission, Hanoi (Viet Nam)

    2004-05-01

    Based on the agreement on the technical cooperation for nuclear technology between Korea and Vietnam, a KAERI/VAEC joint study on the pre-possibility of a new research reactor for Vietnam has been carried out in the research reactor area from Nov. 2003 to May 2004. In this report, the results of the pre-possibility study on a new research reactor are described. The report presents the necessity of a new research reactor in Vietnam, and the desired performance requirements of the new research reactor if necessary. The major design characteristics of some existing research reactors and those under planning were also reviewed and the main characteristics which should be considered in selecting a new multipurpose research reactor for Vietnam were drawn. Some recommendations on the considerations for the next step of the feasibility study such as the project formulation, manpower requirements and international co-operation were also briefly touched upon.

  8. Study on the License Requirements for the SRO/RO of the Kijang Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Subeom; Shin, Taemyung [Korea Nat. University of Transportation, Seoul (Korea, Republic of); Chae, H. T.; Ahn, G. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, S. J.; Gam, S. C. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    The purpose of the study is to propose an appropriate regulatory position for the Kijang reactor operator license requirement by the review of the applicability and compatibility of HANARO SRO/RO license holders for Kijang reactor operation. As the area using radioactive isotope became gradually enlarged both inside and outside of the country, the Kijang research reactor is planned and now under construction next to the HANARO research reactor now being operated in Taejon. In this paper, therefore, an establishment of revised operator license system is discussed for the new research reactor. The design and operation characteristics of the two (HANARO and Kijang) reactors are concluded to be very similar to each other, however, there still exist slight differences in some minor portions. It is recommendable to allow an independent license for each reactor if two reactors of the same power level have recognizable differences in the design and operation characteristics.

  9. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    November 9--10, 1978, marked the first of what has become an annual event--the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and it established the basis for all later meetings. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort. This report provides presentations and discussions of this original meeting. Individual papers have been cataloged separately.

  10. Upgrading of neutron radiography/tomography facility at research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abd El Bar, Waleed; Mongy, Tarek [Atomic Energy Authority, Cairo (Egypt). ETRR-2; Kardjilov, Nikolay [Helmholtz Zentrum Berlin (HZB) for Materials and Energy, Berlin (Germany)

    2014-03-15

    A state-of-the-art neutron tomography imaging system was set up at the neutron radiography beam tube at the Egypt Second Research Reactor (ETRR-2) and was successfully commissioned in 2013. This study presents a set of tomographic experiments that demonstrate a high quality tomographic image formation. A computer technique for data processing and 3D image reconstruction was used to see inside a copy module of an ancient clay article provided by the International Atomic Energy Agency (IAEA). The technique was also able to uncover tomographic imaging details of a mummified fish and provided a high resolution tomographic image of a defective fire valve. (orig.)

  11. Research in nondestructive evaluation techniques for nuclear reactor concrete structures

    Science.gov (United States)

    Clayton, Dwight; Smith, Cyrus

    2014-02-01

    The purpose of the Materials Aging and Degradation (MAaD) Pathway of the Department of Energy's Light Water Reactor Sustainability (LWRS) Program is to develop the scientific basis for understanding and predicting longterm environmental degradation behavior of material in nuclear power plants and to provide data and methods to assess the performance of systems, structures, and components (SSCs) essential to safe and sustained nuclear power plant operations. The understanding of aging-related phenomena and their impacts on SSCs is expected to be a significant issue for any nuclear power plant planning for long-term operations (i.e. service beyond the initial license renewal period). Management of those phenomena and their impacts during long-term operations can be better enable by improved methods and techniques for detection, monitoring, and prediction of SSC degradation. The MAaD Pathway R&D Roadmap for Concrete, "Light Water Reactor Sustainability Nondestructive Evaluation for Concrete Research and Development Roadmap", focused initial research efforts on understanding the recent concrete issues at nuclear power plants and identifying the availability of concrete samples for NDE techniques evaluation and testing. [1] An overview of the research performed by ORNL in these two areas is presented here.

  12. Review of the status of low power research reactors and considerations for its development

    Energy Technology Data Exchange (ETDEWEB)

    Lim, In Cheol; Wu, Sang Ik; Lee, Byung Chul; Ha, Jae Joo [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    At present, 232 research reactors in the world are in operation and two thirds of them have a power less than 1 MW. Many countries have used research reactors as the tools for educating and training students or engineers and for scientific service such as neutron activation analysis. As the introduction of a research reactor is considered a stepping stone for a nuclear power development program, many newcomers are considering having a low power research reactor. The IAEA has continued to provide forums for the exchange of information and experiences regarding low power research reactors. Considering these, the Agency is recently working on the preparation of a guide for the preparation of technical specification possibly for a member state to use when wanting to purchase a low power research reactor. In addition, ANS has stated that special consideration should be given to the continued national support to maintain and expand research and test reactor programs and to the efforts in identifying and addressing the future needs by working toward the development and deployment of next generation nuclear research and training facilities. Thus, more interest will be given to low power research reactors and its role as a facility for education and training. Considering these, the status of low power research reactors was reviewed, and some aspects to be considered in developing a low power research reactor were studied.

  13. Sustainable operations in nuclear research reactors. A bibliographical study

    Energy Technology Data Exchange (ETDEWEB)

    Kibrit, Eduardo; Rodrigues de Aquino, Afonso [Cidade Univ., Sao Paolo (Brazil). Inst. de Pesquisas Energeticas e Nucleares; Marotti de Mello, Adriana [Sao Paolo Univ. (Brazil). Faculdade de Economia; Tromboni de Souza Nascimento, Paulo [Sao Paolo Univ. (Brazil). Faculdade de Economia Administracao e Contabilidade

    2017-10-15

    Sustainability is gaining prominence in the area of operations management. By means of a bibliographical research, we identified in literature sustainable operations carried out by operating organizations of nuclear research reactors. The methodology applied consisted in gathering material, descriptive analysis, selection of analytical categories and evaluation of the material collected. The collection of material was performed by a search made on academic and nuclear databases, with keywords structured for the subject of the research. The collected material was analysed and analytical categories on the theme sustainable operations were established. The evaluation of the collected material resulted in references accepted for the study, classified according to the pre-established analytical categories. The results were significant. From then on, a theoretical review on the topic under study was structured, based on pre-defined analytical categories. Thus, we were able to identify gaps in the literature and propose new studies on the subject.

  14. A scoping analysis of the neotronic design for a new South African research reactor / Vermaak J.I.C.

    OpenAIRE

    Vermaak, Jan Izak Cornelius

    2011-01-01

    Together with many other research reactors around the world, the SAFARI–1 reactor has been classified as an ageing research reactor. In order to continue the provision of the current irradiation services, the operator of the reactor, NECSA, needs to consider the replacement of SAFARI–1 with a new large neutron source, and therefore ultimately a new reactor. A replacement research reactor will have to provide irradiation services that primarily include: radio–isotope producti...

  15. Current Status of Periodic Safety Review of HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minjin; Ahn, Guk-Hoon; Lee, Choong Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    A PSR for a research reactor became a legal requirement as the Nuclear Safety Act was amended and came into effect in 2014. This paper describes the current status and methodology of the first Periodic Safety Review (PSR) of HANARO that is being performed. The legal requirements, work plan, and process of implementing a PSR are described. Because this is the first PSR for a research reactor, it is our understating that the operating organization and regulatory body should communicate well with each other to complete the PSR in a timely manner. The first PSR of HANARO is under way. In order to achieve a successful result, activities of the operation organization such as scheduling, maintaining consistency in input data for review, and reviewing the PSR reports that will require intensive resources should be well planned. This means the operating organization needs to incorporate appropriate measures to ensure the transfer of knowledge and expertise arising from the PSR via a contractor to the operation organization. It is desirable for the Regulatory Body to be involved in all stage of the PSR to prevent any waste of resources and minimize the potential for a reworking of the PSR and the need for an additional assessment and review as recommended by foreign experts.

  16. Neutronics analysis of TRIGA Mark II research reactor

    Directory of Open Access Journals (Sweden)

    Haseebur Rehman

    2018-02-01

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

  17. Analysis of Nigeria research reactor-1 thermal power calibration methods

    Energy Technology Data Exchange (ETDEWEB)

    Agbo, Sunday Arome; Ahmed, Yusuf Aminu; Ewa, Ita Okon; Jibrin, Yahaya [Ahmadu Bello University, Zaria (Nigeria)

    2016-06-15

    This paper analyzes the accuracy of the methods used in calibrating the thermal power of Nigeria Research Reactor-1 (NIRR-1), a low-power miniature neutron source reactor located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. The calibration was performed at three different power levels: low power (3.6 kW), half power (15 kW), and full power (30 kW). Two methods were used in the calibration, namely, slope and heat balance methods. The thermal power obtained by the heat balance method at low power, half power, and full power was 3.7 ± 0.2 kW, 15.2 ± 1.2 kW, and 30.7 ± 2.5 kW, respectively. The thermal power obtained by the slope method at half power and full power was 15.8 ± 0.7 kW and 30.2 ± 1.5 kW, respectively. It was observed that the slope method is more accurate with deviations of 4% and 5% for calibrations at half and full power, respectively, although the linear fit (slope method) on average temperature-rising rates during the thermal power calibration procedure at low power (3.6 kW) is not fitting. As such, the slope method of power calibration is not suitable at lower power for NIRR-1.

  18. Outline of the safety research results, in the power reactor field, fiscal year 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) has promoted the safety research in fiscal year of 1996 according to the Fundamental Research on Safety Research (fiscal year 1996 to 2000) prepared on March, 1996. Here is described on the research results in fiscal year 1996, the first year of the 5 years programme, and whole outline of the fundamental research on safety research, on the power reactor field (whole problems on the new nuclear converter and the fast breeder reactor field and problems relating to the power reactor in the safety for earthquake and probability theoretical safety evaluation field). (G.K.)

  19. Performance Test for Neutron Detector and Associated System using Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seongwoo; Park, Sung Jae; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Oh, Se Hyun [USERS, Daejeon (Korea, Republic of); Shin, Ho Cheol [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system.

  20. Reprocessing of research reactor fuel the Dounreay option

    Energy Technology Data Exchange (ETDEWEB)

    Cartwright, P.

    1997-08-01

    Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storage tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.

  1. Assessment of benefits of research reactors in less developed countries. A case study of the Dalat reactor in Vietnam

    Energy Technology Data Exchange (ETDEWEB)

    Hien, P.D. [Vietnam Atomic Energy Agency, Hanoi (Viet Nam)

    1999-08-01

    The analysis of data on nuclear research reactor (NRR) and socio-economic conditions across countries reveals highly significant relationships of reactor power with GDP and R and D expenditure. The trends revealed can be used as preliminary guides for feasibility assessment of investment in a NRR. Concerning reactor performance, i.e. the number of reactor operation days per year, the covariation with R and D expenditure is most significant, but moderate, implying that there are other controlling factors, e.g. the engagement of country in nuclear power development. Thus, the size of the R and D fund is a most significant indicator to look at in reactor planning. Unfortunately, the lack of adequate R and D funding is a common and chronic problem in less developed countries. As NRR is among the biggest R and D investment in less developed countries, adequate cost benefit assessment is rightfully required. In the case of Vietnam, during 15 years of operation of a 500 kW NRR 2300 Ci of radioisotopes were delivered and 45,000 samples were analysed for multielemental compositions. From a pure financial viewpoint these figures would still be insignificant to justify the investment. However, the impact of the reactor on the technological development seems not to be a matter of pro and cons. The status of reactor utilization and lessons learned are presented and discussed. (author)

  2. IGORR-IV -- Proceedings of the fourth meeting of the International Group on Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbalm, K.F. [comp.

    1995-12-31

    The International Group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Twenty-nine papers were presented in five sessions and written versions of the papers or hard copies of the vugraphs used are published in these proceedings. The five sessions were: (1) Operating Research Reactors and Facility Upgrades; (2) Research Reactors in Design and Construction; (3) ANS Closeout Activities; (4) and (5) Research, Development, and Analysis Results.

  3. Antineutrino emission and gamma background characteristics from a thermal research reactor

    CERN Document Server

    Bui, V M; Fallot, M; Communeau, V; Cormon, S; Estienne, M; Lenoir, M; Peuvrel, N; Shiba, T; Cucoanes, A S; Elnimr, M; Martino, J; Onillon, A; Porta, A; Pronost, G; Remoto, A; Thiolliere, N; Yermia, F; Zakari-Issoufou, A -A

    2016-01-01

    The detailed understanding of the antineutrino emission from research reactors is mandatory for any high sensitivity experiments either for fundamental or applied neutrino physics, as well as a good control of the gamma and neutron backgrounds induced by the reactor operation. In this article, the antineutrino emission associated to a thermal research reactor: the OSIRIS reactor located in Saclay, France, is computed in a first part. The calculation is performed with the summation method, which sums all the contributions of the beta decay branches of the fission products, coupled for the first time with a complete core model of the OSIRIS reactor core. The MCNP Utility for Reactor Evolution code was used, allowing to take into account the contributions of all beta decayers in-core. This calculation is representative of the isotopic contributions to the antineutrino flux which can be found at research reactors with a standard 19.75\\% enrichment in $^{235}$U. In addition, the required off-equilibrium correction...

  4. Very high flux research reactors based on particle fuels

    Energy Technology Data Exchange (ETDEWEB)

    Powell, J.R.; Takahashi, H.

    1985-01-01

    A new approach to high flux research reactors is described, the VHFR (Very High Flux Reactor). The VHFR fuel region(s) are packed beds of HTGR-type fuel particles through which coolant (e.g., D/sub 2/O) flows directly. The small particle diameter (typically on the order of 500 microns) results in very large surface areas for heat transfer (approx. 100 cm/sup 2//cm/sup 3/ of bed), high power densities (approx. 10 megawatts per liter), and minimal ..delta..T between fuel and coolant (approx. 10 K) VHFR designs are presented which achieve steady-state fluxes of approx. 2x10/sup 16/ n/cm/sup 2/sec. Deuterium/beryllium combinations give the highest flux levels. Critical mass is low, approx. 2 kg /sup 235/U for 20% enriched fuel. Refueling can be carried out continuously on-line, or in a batch process with a short daily shutdown. Fission product inventory is very low, approx. 100 to 300 grams, depending on design.

  5. Progress report on neutron beam experiments at Dalat Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen Canh Hai; Tran Tuan Anh [Nuclear Physics Department, Nuclear Research Institute, Dalat (Viet Nam)

    2000-10-01

    The conduct and the utilizations of neutron beams at Dalat Nuclear Research Reactor was reported. In 1998 and 1999 the filtered thermal neutron beam at the beam tube using substances Si, Ti, C, Pb was extracted. The investigations on physical characteristics of reactor; neutron spectra and fluxes at beam tube; safety conditions have been carried out by calculations and experiments. The physical characteristics for the purposes of prompt gamma neutron activation analysis (PGNAA) and nuclear data measurement were improved. The delayed neutron analysis method is used to detect the neutron emission fragments produced by neutron irradiation of uranium and thorium. This method is to determine the concentrations of uranium and thorium simultaneously and detect 10{sup -6} g of uranium and 8x10{sup -6} g thorium in 10 g sample with a precision of 8 per cent. Beside the delayed neutron analysis facility, a gamma spectrometer system with HP-Ge 90 cm{sup 3} semiconductor detector was installed at pneumatic transfer for cyclic activation analysis (CAA). The CAA method has given analytical results quickly and sensitively for the isotopes with half-lives in order of seconds to minutes. (author)

  6. Ageing investigation and upgrading of components/systems of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip; Widi Setiawan [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia)

    1998-10-01

    Kartini research reactor has been operated in good condition and has demonstrated successful operation for the past 18 years, utilized for: reactor kinetic and control studies, instrumentation tests, neutronic and thermohydraulic studies, routine neutron activation analysis, reactor safety studies, training for research reactor operators and supervisors, and reactor physics experiments. Several components of Kartini reactor use components from the abandoned IRT-2000 Project at Serpong and from Bandung Reactor Centre such as: reactor tank, reactor core, heat exchanger, motor blower for ventilation system, fuel elements, etc. To maintain a good operating performance and also for aging investigation purposes, the component failure data collection has been done. The method used is based on the Manual on Reliability Data Collection For Research Reactor PSAs, IAEA TECDOC 636, and analyzed by using Data Entry System (DES) computer code. Analysis result shows that the components/systems failure rate of Kartini reactor is around 1,5.10{sup -4} up to 2,8.10{sup -4} per hour, these values are within the ranges of the values indicated in IAEA TECDOC 478. Whereas from the analysis of irradiation history shows that the neutron fluence of fuel element with highest burn-up (2,05 gram U-235 in average) is around 1.04.10{sup 16} n Cm{sup -2} and this value is still far below its limiting value. Some reactor components/systems have been replaced and upgraded such as heat exchanger, instrumentation and control system (ICS), etc. The new reactor ICS was installed in 1994 which is designed as a distributed structure by using microprocessor based systems and bus system technology. The characteristic and operating performance of the new reactor ICS, as well as the operation history and improvement of the Kartini research reactor is presented. (J.P.N.)

  7. An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rosenthal, Murray Wilford [ORNL

    2009-08-01

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

  8. Sodium fast reactor fuels and materials : research needs.

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R.; Porter, Douglas (Idaho National Laboratory, Idaho Falls, ID); Wright, Art (Argonne National Laboratory Argonne, IL); Lambert, John (Argonne National Laboratory Argonne, IL); Hayes, Steven (Idaho National Laboratory, Idaho Falls, ID); Natesan, Ken (Argonne National Laboratory Argonne, IL); Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Garner, Frank (Radiation Effects Consulting. Richland, WA); Walters, Leon (Advanced Reactor Concepts, Idaho Falls, ID); Yacout, Abdellatif (Argonne National Laboratory Argonne, IL)

    2011-09-01

    An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

  9. Issues and future direction of thermal-hydraulics research and development in nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Saha, P., E-mail: pradip.saha@ge.com [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Aksan, N. [GRNSPG Group, University of Pisa (Italy); Andersen, J. [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Yan, J. [Westinghouse Electric Co., Columbia, SC (United States); Simoneau, J.P. [AREVA, Lyon (France); Leung, L. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada); Bertrand, F. [CEA, DEN, DER, F-13108 Saint-Paul-Lez-Durance (France); Aoto, K.; Kamide, H. [Japan Atomic Energy Agency, Chiyoda-ku, Tokyo (Japan)

    2013-11-15

    The paper archives the proceedings of an expert panel discussion on the issues and future direction of thermal-hydraulic research and development in nuclear power reactors held at the NURETH-14 conference in Toronto, Canada, in September 2011. Thermal-hydraulic issues related to both operating and advanced reactors are presented. Advances in thermal-hydraulics have significantly improved the performance of operating reactors. Further thermal-hydraulics research and development is continuing in both experimental and computational areas for operating reactors, reactors under construction or ready for near-term deployment, and advanced Generation-IV reactors. As the computing power increases, the fine-scale multi-physics computational models, coupled with the systems analysis code, are expected to provide answers to many challenging problems in both operating and advanced reactor designs.

  10. Analysis of Nigeria Research Reactor-1 Thermal Power Calibration Methods

    Directory of Open Access Journals (Sweden)

    Sunday Arome Agbo

    2016-06-01

    Full Text Available This paper analyzes the accuracy of the methods used in calibrating the thermal power of Nigeria Research Reactor-1 (NIRR-1, a low-power miniature neutron source reactor located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. The calibration was performed at three different power levels: low power (3.6 kW, half power (15 kW, and full power (30 kW. Two methods were used in the calibration, namely, slope and heat balance methods. The thermal power obtained by the heat balance method at low power, half power, and full power was 3.7 ± 0.2 kW, 15.2 ± 1.2 kW, and 30.7 ± 2.5 kW, respectively. The thermal power obtained by the slope method at half power and full power was 15.8 ± 0.7 kW and 30.2 ± 1.5 kW, respectively. It was observed that the slope method is more accurate with deviations of 4% and 5% for calibrations at half and full power, respectively, although the linear fit (slope method on average temperature-rising rates during the thermal power calibration procedure at low power (3.6 kW is not fitting. As such, the slope method of power calibration is not suitable at lower power for NIRR-1.

  11. Diversion assumptions for high-powered research reactors. ISPO C-50 Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    Binford, F.T.

    1984-01-01

    This study deals with diversion assumptions for high-powered research reactors -- specifically, MTR fuel; pool- or tank-type research reactors with light-water moderator; and water, beryllium, or graphite reflectors, and which have a power level of 25 MW(t) or more. The objective is to provide assistance to the IAEA in documentation of criteria and inspection observables related to undeclared plutonium production in the reactors described above, including: criteria for undeclared plutonium production, necessary design information for implementation of these criteria, verification guidelines including neutron physics and heat transfer, and safeguards measures to facilitate the detection of undeclared plutonium production at large research reactors.

  12. Sodium fast reactor safety and licensing research plan. Volume I.

    Energy Technology Data Exchange (ETDEWEB)

    Sofu, Tanju (Argonne National Laboratory, Argonne, IL); LaChance, Jeffrey L.; Bari, R. (Brokhaven National Laboratory Upton, NY); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.; Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN)

    2012-05-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  13. SoLid: Search for Oscillation with a 6Li Detector at the BR2 research reactor

    OpenAIRE

    Michiels, Ianthe

    2016-01-01

    In the past decades, various nuclear reactor neutrino experiments have measured a deficit in the flux of antineutrinos coming from the reactor at short reactor-detector distances, when compared to theoretical calculations. One of the experiments designed to investigate this reactor antineutrino anomaly is the SoLid experiment. It uses the compact BR2 research reactor from the SCK-CEN in Mol, Belgium, to perform reactor antineutrino flux measurements at very short baseline. These proceedings d...

  14. Research and proposal on SCR reactor optimization for industrial boiler.

    Science.gov (United States)

    Yang, Yiming; Li, Jian; He, Hong

    2017-08-24

    The advanced CFD software STAR-CCM+ was used to simulate a denitrification (De-NOx) project for a boiler in this paper, and the simulation result was verified based on a physical model. Two SCR reactors were developed: reactor #1 was optimized and #2 was developed based on #1. Various indicators including gas flow field, ammonia concentration distribution, temperature distribution, gas incident angle and system pressure drop were analyzed. The analysis indicated Reactor #2 was of outstanding performance and could simplify developing greatly. Ammonia injection grid (AIG), the core component of reactor was studied; three AIGs were developed and their performances were compared and analyzed. The result indicated that AIG #3 was of the best performance. The technical indicators were proposed for SCR reactor based on the study. Flow filed distribution, gas incident angle and temperature distribution are subjected to SCR reactor shape to a great extent and Reactor #2 proposed in this paper was of outstanding performance; ammonia concentration distribution is subjected to Ammonia injection grid (AIG) shape and AIG #3 could meet the technical indicator of ammonia concentration without mounting ammonia mixer. The development above on the reactor and the AIG are both of great application value and social efficiency.

  15. 75 FR 27368 - Aerotest Operations, Inc., Aerotest Radiography and Research Reactor; Notice of Consideration of...

    Science.gov (United States)

    2010-05-14

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Aerotest Operations, Inc., Aerotest Radiography and Research Reactor; Notice of Consideration of... Operating License No. R-98 for the Aerotest Radiography and Research Reactor (ARRR), currently held by...

  16. 75 FR 39985 - In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order...

    Science.gov (United States)

    2010-07-13

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order..., use and operation of the Aerotest Radiography and Research Reactor (ARRR) located in San Ramon...

  17. Implementation of a management system for operating organizations of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kibrit, Eduardo, E-mail: kibrit@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Aquino, Afonso Rodrigues de; Zouain, Desiree Moraes, E-mail: araquino@ipen.b, E-mail: dmzouain@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This paper presents the requirements established by an IAEA draft technical document for the implementation of a management system for operating organisations of research reactors. The following aspects will be discussed: structure of IAEA draft technical document, management system requirements, processes common to all research reactors, aspects for the implementation of the management system, and a formula for grading the management system requirements. (author)

  18. Status of reduced enrichment programs for research reactors in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kanda, Keiji; Nishihara, Hedeaki [Kyoto Univ., Osaka (Japan); Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo [Japan Atomic Energy Research Institute, Tokyo (Japan)

    1997-08-01

    The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for the full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE.

  19. Monitoring and Control Research Using a University Reactor and SBWR Test-Loop

    Energy Technology Data Exchange (ETDEWEB)

    Robert M. Edwards

    2003-09-28

    The existing hybrid simulation capability of the Penn State Breazeale nuclear reactor was expanded to conduct research for monitoring, operations and control. Hybrid simulation in this context refers to the use of the physical time response of the research reactor as an input signal to a real-time simulation of power-reactor thermal-hydraulics which in-turn provides a feedback signal to the reactor through positioning of an experimental changeable reactivity device. An ECRD is an aluminum tube containing an absorber material that is positioned in the central themble of the reactor kinetics were used to expand the hybrid reactor simulation (HRS) capability to include out-of-phase stability characteristics observed in operating BWRs.

  20. Impact of a 2 MWth research reactor on radioactivity in sediments.

    Science.gov (United States)

    Bose, S R; Williamson, T G; Mulder, R U; Molla, M A

    1993-08-01

    The objective of this study was to determine the environmental impact caused by liquid effluent discharge from the University of Virginia's 2 MWth Research Reactor during the course of its first quarter-century of normal operation. Sediment samples were collected from the reactor pond (situated in a restricted area next to the Reactor Facility), the stream feeding it, and its exit stream. For a comparative study, sediment samples were taken from a nearby closed reference pond having no direct link with the reactor pond. Concentrations of long-lived alpha, beta and gamma emitting radionuclides, from natural and nuclear weapons fallout sources, were detected in pond and stream sediments. Low levels of activation product radioisotopes from the research reactor were detected in the reactor pond sediment. It was observed that both natural and artificial radionuclide concentrations were higher in the UVAR pond (with the exception of 54Mn and 65Zn) as compared to exist stream and reference sediments.

  1. Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

    Energy Technology Data Exchange (ETDEWEB)

    Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.; Holdaway, K. K.; Housley, G. K.; Rabin, B. H.

    2016-10-01

    The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, and other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.

  2. Research about reactor operator's personability characteristics and performance

    Energy Technology Data Exchange (ETDEWEB)

    Wei Li; He Xuhong; Zhao Bingquan [Tsinghua Univ., Institute of Nuclear Energy Technology, Beijing (China)

    2003-03-01

    To predict and evaluate the reactor operator's performance by personality characteristics is an important part of reactor operator safety assessment. Using related psychological theory combined with the Chinese operator's fact and considering the effect of environmental factors to personality analysis, paper does the research about the about the relationships between reactor operator's performance and personality characteristics, and offers the reference for operator's selection, using and performance in the future. (author)

  3. Future development of the research nuclear reactor IRT-2000 in Sofia

    Energy Technology Data Exchange (ETDEWEB)

    Apostolov, T.G. [Institute for Nuclear Research and Nuclear Energy, BAS, Sofia (Bulgaria)

    1999-07-01

    The present paper presents a short description of the research reactor IRT-2000 Sofia, started in 1961 and operated for 28 years. Some items are considered, connected to the improvements made in the contemporary safety requirements and the unrealized project for modernization to 5 MW. Proposals are considered for reconstruction of reactor site to a 'reactor of low power' for education purposes and as a basis for the country's nuclear technology development. (author)

  4. U.S. Department of Energy Program of International Technical Cooperation for Research Reactor Utilization

    Energy Technology Data Exchange (ETDEWEB)

    Chong, D.; Manning, M.; Ellis, R.; Apt, K.; Flaim, S.; Sylvester, K.

    2004-10-03

    The U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) has initiated collaborations with the national nuclear authorities of Egypt, Peru, and Romania for the purpose of advancing the commercial potential and utilization of their respective research reactors. Under its Office of International Safeguards ''Sister Laboratory'' program, DOE/NNSA has undertaken numerous technical collaborations over the past decade intended to promote peaceful applications of nuclear technology. Among these has been technical assistance in research reactor applications, such as neutron activation analysis, nuclear analysis, reactor physics, and medical radioisotope production. The current collaborations are intended to provide the subject countries with a methodology for greater commercialization of research reactor products and services. Our primary goal is the transfer of knowledge, both in administrative and technical issues, needed for the establishment of an effective business plan and utilization strategy for the continued operation of the countries' research reactors. Technical consultation, cooperation, and the information transfer provided are related to: identification, evaluation, and assessment of current research reactor capabilities for products and services; identification of opportunities for technical upgrades for new or expanded products and services; advice and consultation on research reactor upgrades and technical modifications; characterization of markets for reactor products and services; identification of competition and estimation of potential for market penetration; integration of technical constraints; estimation of cash flow streams; and case studies.

  5. Nuclear plant-aging research on reactor protection systems

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, L.C.

    1988-01-01

    This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed.

  6. Radioactive waste management in Research Reactor Institute, Kyoto University

    Energy Technology Data Exchange (ETDEWEB)

    Shimoura, Kazukuni [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.

    1997-02-01

    The joint utilization by the researchers of the universities and others in whole Japan of the reactor facilities in Kyoto University was begun in 1965. The facility for abandoning radioactive waste was constructed in fiscal year 1963, and comprises 500 kg/h vaporization and concentration facility, 2 systems of 5 m{sup 3}/h flocculation, precipitation and filtration facility, and 2 systems of 5 m{sup 3}/h ion exchange facility for liquid waste, 50x10{sup 3} kg pressing capacity, four-column type press for reducing volume of solid waste, and waste store for 400 standard drums. Radioactive Waste Treatment Department was organized, and the stipulations on security and radiation injury prevention were enforced in 1964. Liquid and solid wastes have been accepted since 1964. The radioactivity in discharged water in each year is shown. About 600 m{sup 3} of waste liquid has been accepted in one year since 1980. The trust of solid waste treatment to Japan Radioisotope Association has been carried out 51 times. The radioactive waste which is temporarily stored in the waste store is reported. Hereafter, the construction of the facility for storing large finished equipment and the appearance of waste treatment enterprises are desirable. (K.I.)

  7. High resolution neutron diffractometer HRND at research reactor CMRR

    Science.gov (United States)

    Zhang, J.; Xia, Y.; Wang, Y.; Xie, C.; Sun, G.; Liu, L.; Pang, B.; Li, J.; Huang, C.; Liu, Y.; Gong, J.

    2018-01-01

    The high resolution neutron diffractometer HRND is located at the 20 MW China Mianyang Research Reactor (CMRR), which is a neutron powder diffractometer especially dedicated to crystal and magnetic structure studies for polycrystalline powder samples. A vertical focusing Ge (511) monochromator produce a monochromatic neutron beam with a wavelength of 1.885 Å at a fixed take-off angle of 120o. An array of 64 equidistant 3He filled proportional counters can acquire diffraction patterns with a large-scale diffraction angle range over 160o. As all the Soller slit collimators of HRND have a collimation angle of 10' and the monochromator has an average mosaicity of 0.359o, HRND obtains a best resolution of about 1.6\\textperthousand based on experiments, which makes the resolution of HRND can compete with the mainstream-level high resolution neutron powder diffractometers in the world. Equipped with a cryostat and a furnace, HRND allows structural characterization in an extremely broad temperature range. The details of the configuration and performance of the instrument are reported along with its specifications and performance assessments in the present paper.

  8. A neutron tomography facility at a low power research reactor

    CERN Document Server

    Körner, S; Von Tobel, P; Rauch, H

    2001-01-01

    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at ...

  9. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  10. A Study on the demands of research reactors and considerations for an export

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, Young Jun

    2008-11-15

    Among around 240 research reactors in operation over the world, around 80% have been operated for more than 20 years and 65% for more than 30 years. Hence the number of operable reactors is expected, between 2010 and 2020, to be reduced to 1/3 of the present situation if the lifetime of a research reactor is assumed to be 40 years. However, considering the recent re-highlighting of nuclear energy as a practical mass energy source and the contributions to the overall areas of science and technology, the demands for constructing a new research reactor and replacing the existing research reactors will be increased in the near future. On the other hand, vendors which participate in providing research reactors are not few, and AREVA in France and INVAP in Argentina are example of them in a positive position. Japan and Russia are regarded as potential competitors, but they do not actively appear in the market so far. Comparing those competitors with Korea, we have weak points regarding experiences on exports and the organizational systems as an integrated vendor. But we may have a competitiveness by grafting our experiences on the development of nuclear power technology and the construction and operation of the HANARO. In this report, the future potential demands for research reactors and the related considerations for exports have been surveyed and described, particularly, centering around the Netherlands, Vietnam and Thailand that are countries which may construct research reactors in the near future. Considerations for exporting a research reactor have been categorized into two groups of technical and nontechnical items. From a technical point of view, the issues on fuel and reactor type, design data and design ability, design codes, and technology property rights have been reviewed. For the non-technical items, an integrated project system, reasonable estimate of demands, social and economic conditions for potential demand countries, MOU status, nuclear non

  11. Status of reactor-shielding research in the US

    Energy Technology Data Exchange (ETDEWEB)

    Maienshein, F.C.

    1980-01-01

    While reactor programs change, shielding analysis methods are improved slowly. Version-V of ENDF/B provides improved data and Version-VI will be cost effective in advanced fission reactors are to be developed in the US. Benchmarks for data and methods validation are collected and distributed in the US in two series, one primarily for FBR-related experiments and one for LWR calculational methods. For LWR design, cavity streaming is now handled adequately, if with varying degrees of elegance. Investigations of improved detector response for LWRs rely upon transport methods. The great potential importance of pressure-vessel damage is dreflected in widespread studies to aid in the prediction of neutron fluences in vessels. For LMFBRS, the FFTF should give attenuation results on an operating reactor. For larger power reactors, the advantages of alternate shield materials appear compelling. A few other shielding studies appear to require experimental confirmation if LMFBRs are to be economically competitive. A coherent shielding program for the GCFR is nearing completion. For the fusion-reactor program, methods verification is under way, practical calculations are well advanced for test devices such as the TFTR and FMIT, and consideration is now given to shielding problems of large reactors, as in the ETF study.

  12. Reactor Safety Research Programs Quarterly Report October - December 1981

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1982-03-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from October 1 through December 31, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where serviceinduced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and post accident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  13. Reactor Safety Research Programs Quarterly Report July - September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1982-01-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from July 1 through September 30, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR} steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  14. Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharyya, S. K.; Boing, L. E.

    2000-02-17

    The aging of research reactors worldwide has resulted in a heightened awareness in the international technical decommissioning community of the timeliness to review and address the needs of these research institutes in planning for and eventually performing the decommissioning of these facilities. By using the reactors already undergoing decommissioning as test beds for evaluating enhanced or new/innovative technologies for decommissioning, it is possible that new techniques could be made available for those future research reactor decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the research institutes in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research reactor decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to use in the research reactors. The decommissioning of the CP-5 Research Reactor is currently in the final phase of dismantlement. In this paper the authors present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors.

  15. Burnup measurements on spent fuel elements of the RP-10 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

  16. Nuclear energy was the way of the future; 50 anniversary of the research reactor

    NARCIS (Netherlands)

    Wassink, J.

    2013-01-01

    It was the hidden jewel of TU Delft, according to the employees of the nuclear reactor. Others protested against it and insisted that it be eliminated. Following a major mid-life crisis, the Delft research reactor is now in better shape than ever before.

  17. Generating the flux map of Nigeria Research Reactor-1 for efficient ...

    African Journals Online (AJOL)

    One of the main uses to which the Nigeria Research Reactor-1 (NIRR-1) will be put is neutron activation analysis. The activation analyst requires information about the flux level at various points within and around the reactor core to enable him identify the point of optimum flux (at a given operating power) for any irradiation ...

  18. Conceptual Nuclear Design of a 20 MW Multipurpose Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Kim, Hak Sung; Park, Cheol [KAERI, Daejeon (Korea, Republic of); Nghiem, Huynh Ton; Vinh, Le Vinh; Dang, Vo Doan Hai [Dalat Nuclear Research Reactor, Hanoi (Viet Nam)

    2007-08-15

    A conceptual nuclear design of a 20 MW multi-purpose research reactor for Vietnam has been jointly done by the KAERI and the DNRI (VAEC). The AHR reference core in this report is a right water cooled and a heavy water reflected open-tank-in-pool type multipurpose research reactor with 20 MW. The rod type fuel of a dispersed U{sub 3}Si{sub 2}-Al with a density of 4.0 gU/cc is used as a fuel. The core consists of fourteen 36-element assemblies, four 18-element assemblies and has three in-core irradiation sites. The reflector tank filled with heavy water surrounds the core and provides rooms for various irradiation holes. Major analyses have been done for the relevant nuclear design parameters such as the neutron flux and power distributions, reactivity coefficients, control rod worths, etc. For the analysis, the MCNP, MVP, and HELIOS codes were used by KAERI and DNRI (VAEC). The results by MCNP (KAERI) and MVP (DNRI) showed good agreements and can be summarized as followings. For a clean, unperturbed core condition such that the fuels are all fresh and there are no irradiation holes in the reflector region, the fast neutron flux (E{sub n}{>=}1.0 MeV) reaches 1.47x10{sup 14} n/cm{sup 2}s and the maximum thermal neutron flux (E{sub n}{<=}0.625 eV) reaches 4.43x10{sup 14} n/cm{sup 2}s in the core region. In the reflector region, the thermal neutron peak occurs about 28 cm far from the core center and the maximum thermal neutron flux is estimated to be 4.09x10{sup 14} n/cm{sup 2}s. For the analysis of the equilibrium cycle core, the irradiation facilities in the reflector region were considered. The cycle length was estimated as 38 days long with a refueling scheme of replacing three 36-element fuel assemblies or replacing two 36-element and one 18-element fuel assemblies. The excess reactivity at a BOC was 103.4 mk, and 24.6 mk at a minimum was reserved at an EOC. The assembly average discharge burnup was 54.6% of initial U-235 loading. For the proposed fuel management

  19. 78 FR 26811 - Dow Chemical Company, Dow TRIGA Research Reactor; License Renewal for the Dow Chemical TRIGA...

    Science.gov (United States)

    2013-05-08

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

  20. Digital, remote control system for a 2-MW research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Battle, R.E.; Corbett, G.K.

    1988-01-01

    A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs.

  1. Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor.

    Science.gov (United States)

    Ajijul Hoq, M; Malek Soner, M A; Salam, M A; Haque, M M; Khanom, Salma; Fahad, S M

    2017-12-01

    The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

  2. Multipurpose epithermal neutron beam on new research station at MARIA research reactor in Swierk-Poland

    Energy Technology Data Exchange (ETDEWEB)

    Gryzinski, M.A.; Maciak, M. [National Centre for Nuclear Research, Andrzeja Soltana 7, 05-400 Otwock-Swierk (Poland)

    2015-07-01

    MARIA reactor is an open-pool research reactor what gives the chance to install uranium fission converter on the periphery of the core. It could be installed far enough not to induce reactivity of the core but close enough to produce high flux of fast neutrons. Special design of the converter is now under construction. It is planned to set the research stand based on such uranium converter in the near future: in 2015 MARIA reactor infrastructure should be ready (preparation started in 2013), in 2016 the neutron beam starts and in 2017 opening the stand for material and biological research or for medical training concerning BNCT. Unused for many years, horizontal channel number H2 at MARIA research rector in Poland, is going to be prepared as a part of unique stand. The characteristics of the neutron beam will be significant advantage of the facility. High flux of neutrons at the level of 2x10{sup 9} cm{sup -2}s{sup -1} will be obtainable by uranium neutron converter located 90 cm far from the reactor core fuel elements (still inside reactor core basket between so called core reflectors). Due to reaction of core neutrons with converter U{sub 3}Si{sub 2} material it will produce high flux of fast neutrons. After conversion neutrons will be collimated and moderated in the channel by special set of filters and moderators. At the end of H2 channel i.e. at the entrance to the research room neutron energy will be in the epithermal energy range with neutron intensity at least at the level required for BNCT (2x10{sup 9} cm{sup -2}s{sup -1}). For other purposes density of the neutron flux could be smaller. The possibility to change type and amount of installed filters/moderators which enables getting different properties of the beam (neutron energy spectrum, neutron-gamma ratio and beam profile and shape) is taken into account. H2 channel is located in separate room which is adjacent to two other empty rooms under the preparation for research laboratories (200 m2). It is

  3. Progress with OPAL, the new Australian research reactor

    Indian Academy of Sciences (India)

    Abstract. Australian science is entering a new 'golden age', with the start-up of bright new neutron and photon sources in Sydney and Melbourne, in 2006 and 2007 respectively. The OPAL reactor and the Australian Synchrotron can be considered as the greatest sin- gle investment in scientific infrastructure in Australia's ...

  4. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Administrator

    Cs, will be separated and used as a radiation source for various societal applications. This approach minimizes the quantity of waste to be immobilized. Separation of noble metals such as palladium for societal applications such as catalysts, fuel cells etc is also possible. 3. FBR Programme in India. The seed for fast reactor ...

  5. IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra; Filho, Walter Ricci [Nuclear and Energy Research Institute, IPEN-CNEN/SP, Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242 Cid Universitaria CEP: 05508-000- Sao Paulo-SP (Brazil)

    2015-07-01

    IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors. The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)

  6. 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)

  7. Water chemistry management of research reactor in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yoshijima, Tetsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    The JRR-3M cooling system consists of four systems, namely; (1) primary cooling system, (2) heavy water cooling system, (3) helium system and (4) secondary cooling system. The heavy water is used for reflector and pressurized with helium gas. Water chemistry management of the JRR-3M cooling systems is one of the important subject for the safety operation. The main objects are to prevent the corrosion of cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the generation of radioactive waste. All measured values were within the limits of specifications and JRR-3M reactor was operated with safety in 1996. Spent fuels of JRR-3M reactor are stored in the spent fuel pool. This pool water has been analyzed to prevent corrosion of aluminum cladding of spent fuels. Water chemistry of spent fuel pool water is applied to the prevention of corrosion of aluminum alloys including fuel cladding. The JRR-2 reactor was eternally stopped in December 1996 and is now under decommissioning. The JRR-2 reactor is composed of heavy water tank, fuel guide tube and horizontal experimental hole. These are constructed of aluminum alloy and biological shield and upper shield are constructed of concrete. Three types of corrosion of aluminum alloy were observed in the JRR-2. The Alkaline corrosion of aluminum tube occurred in 1972 because of the mechanical damage of the aluminum fuel guide tube which is used for fuel handling. Modification of the reactor top shield was started in 1974 and completed in 1975. (author)

  8. Proceedings of the 1997 workshop on the utilization of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-10-01

    The 1997 Workshop on the Utilization of Research Reactors, which is the sixth Workshop on the theme of research reactor utilization was held in Bandung, Indonesia from November 6 to 13. This Workshop was executed based on the agreement in the Eighth International conference for Nuclear Cooperation in Asia (ICNCA) held in Tokyo, March 1997. The whole Workshop consists of the preceding Sub-workshop carried out the demonstration experiment of Radioisotope Production, and the Workshop on the theme of three fields (Neutron Scattering, Radioisotope production, Safe Operation and Maintenance of Research Reactor). The total number of participants for the workshop was about 100 people from 8 countries, i.e. China, Indonesia, Korea, Malaysia, Philippine, Thailand, Vietnam and Japan. It consists of the papers for Sub-workshop, Neutron Scattering, Radioisotope Production, Safe Operation and Maintenance of research reactor, and summary reports. The 53 of the presented papers are indexed individually. (J.P.N.)

  9. Improvement of Critical Heat Flux Correlation for Research Reactors using Plate-Type Fuel

    National Research Council Canada - National Science Library

    KAMINAGA, Masanori; YAMAMOTO, Kazuyoshi; SUDO, Yukio

    1998-01-01

    ... reversal.The authors have proposed a CHF correlation scheme for the thermal hydraulic design of research reactors, based on CHF experiments for both upward and downward flows including CCFL condition...

  10. Review of the nuclear reactor thermal hydraulic research in ocean motions

    Energy Technology Data Exchange (ETDEWEB)

    Yan, B.H., E-mail: yanbh3@mail.sysu.edu.cn

    2017-03-15

    The research and development of small modular reactor in floating platform has been strongly supported by Chinese government and enterprises. Due to the effect of ocean waves, the thermal hydraulic behavior and safety characteristics of floating reactor are different from that of land-based reactor. Many scholars including the author have published their research and results in open literatures. Much of these literatures are valuable but there are also some contradictory conclusions. In this wok, the nuclear reactor thermal hydraulic research in ocean motions was systematically summarized. Valuable results and experimental data were analyzed and classified. Inherent mechanism for controversial issues in different experiments was explained. Necessary work needed in the future was suggested. Through this work, we attempt to find as many valuable results as possible for the designing and subsequent research.

  11. A pragmatic approach towards designing a second shutdown system for Tehran research reactor

    National Research Council Canada - National Science Library

    Boustani Ehsan; Khakshournia Samad; Khalafi Hossein

    2016-01-01

    One second shutdown system is proposed for the Tehran Research Reactor to achieve the goal of higher safety in compliance with current operational requirements and regulations and improve the overall...

  12. Core calculations for the upgrading of the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: asantos@net.ipen.br; perrotta@net.ipen.br; mitsuo@net.ipen.br

    1998-07-01

    The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)

  13. Nuclear materials testing in the loops of the NRU research reactor using material test bundles

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C.; Walters, L. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2014-07-01

    The NRU research reactor has been used to obtain data to understand and quantify the effects of irradiation on nuclear reactor components through their in-service lives and to develop improved designs and components. Apart from the Mark-4 and Mark-7 fast neutron rod material testing facilities in NRU, the high-pressure/high-temperature experimental loops provide an environment similar to the CANDU reactor core, where test materials are subjected to simulated power reactor conditions. Nuclear materials are tested in the loops using Material Test Bundles (MTB). This paper describes how the MTB is designed to operate in the NRU loops. It also describes the physics calculation of the 89-energy-group neutron spectrum in the MTB and its comparison with the spectrum in CANDU power reactors. The predictions of spectral effects on nuclear material behaviour, such as material damage and helium generation are summarized. (author)

  14. Design and installation of a hot water layer system at the Tehran research reactor

    Directory of Open Access Journals (Sweden)

    Mirmohammadi Sayedeh Leila

    2013-01-01

    Full Text Available A hot water layer system (HWLS is a novel system for reducing radioactivity under research reactor containment. This system is particularly useful in pool-type research reactors or other light water reactors with an open pool surface. The main purpose of a HWLS is to provide more protection for operators and reactor personnel against undesired doses due to the radio- activity of the primary loop. This radioactivity originates mainly from the induced radioactivity contained within the cooling water or probable minute leaks of fuel elements. More importantly, the bothersome radioactivity is progressively proportional to reactor power and, thus, the HWLS is a partial solution for mitigating such problems when power upgrading is planned. Following a series of tests and checks for different parameters, a HWLS has been built and put into operation at the Tehran research reactor in 2009. It underwent a series of comprehensive tests for a period of 6 months. Within this time-frame, it was realized that the HWLS could provide a better protection for reactor personnel against prevailing radiation under containment. The system is especially suitable in cases of abnormality, e. g. the spread of fission products due to fuel failure, because it prevents the mixing of pollutants developed deep in the pool with the upper layer and thus mitigates widespread leakage of radioactivity.

  15. Proceedings of the first symposium on utilization of research reactors and JMTR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The first symposium on utilization of research reactors (JRR-2, JRR-3M, JRR-4) and Japan Materials Testing Reactor (JMTR) in JAERI was held from September 29th to 30th, 1997 at Sannomaru Hotel, Mito. The purpose of this symposium is to announce contribution to progress of scientific technology as well as to promote future utilization of the research reactors and JMTR. During the symposium, 16 reports were presented on nuclear fuel and material, neutron beam experiment, medical irradiation, radioisotope production and neutron activation analysis. The present status of the research reactors and JMTR were also reported. The special lecture titled `JRR-2 and Medical Irradiation` was given by Mr. Nakamura, former editorial writer of Yomiuri. Finally, panel discussion was carried on `The Role of Research Reactors and JMTR in Scientific Technology for the future` actively by the participants and experts in every field of research reactor utilization. 250 people participated in this symposium from universities, national research institutes, private corporations and JAERI. This proceedings briefly summarizes 16 reports, the content of panel discussion and so forth. (J.P.N.)

  16. Research reactor systems for the stable and efficient supply of RI

    Energy Technology Data Exchange (ETDEWEB)

    Lim, In Choel; Oh, Sooy Oul; Lee, Choong Sung; Jun, Byung Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-04-15

    The supply of medical isotopes has relied on the major four research reactors over the world and an unavailability of some of these reactors caused a problem in the stable supply of medical isotopes, especially {sup 9}9{sup M}o. There are several ways to produce {sup 9}9{sup M}o but is is believed that the use of a research reactor is the most efficient way. There are two ways to produce {sup 9}9{sup M}o in a research reactor; they are the separation of {sup 9}9{sup M}o from the fission product and the use of neutron capture reaction of {sup 9}8{sup M}o. For the former, various ways are available depending on the target morphology and the enrichment of uranium in the target. The efficiency of the neutron capture method depends on the available neutron flux, the enrichment of {sup 9}8{sup M}o in the target and the efficiency of the adsorption column. Besides these nuclear engineering aspects, other issues affect the use of the research reactor and they include the following; the on power loading of the target, the methods to reduce the cost for the production of RI in research reactors, the logistics between the producer and the consumer, and the coalition of research reactors. In addition, the producers of RI products or the distributors should become the prosumers in the production of sources. The stable and efficient supply of medical isotopes is believed to depend on all these factors and the future options on the use of a research reactor in Korea for the medical isotope supply should consider these.

  17. ITHNA.SYS: An Integrated Thermal Hydraulic and Neutronic Analyzer SYStem for NUR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mazidi, S., E-mail: samirmazidi@gmail.com [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Meftah, B., E-mail: b_meftah@yahoo.com [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Belgaid, M., E-mail: belgaidm@yahoo.com [Faculté de Physique, Université Houari Boumediene, USTHB, BP 31, Bab Ezzouar, Alger (Algeria); Letaim, F., E-mail: fletaim@yahoo.fr [Faculté des Sciences et Technologies, Université d’El-oued, PO Box 789, El-oued (Algeria); Halilou, A., E-mail: hal_rane@yahoo.fr [Division Réacteur NUR, Centre de Recherche Nucléaire de Draria, BP 43 Sebala, Draria, Alger (Algeria)

    2015-08-15

    Highlights: • We develop a neutronic and thermal hydraulic MTR reactor analyzer. • The analyzer allows a rapid determination of the reactor core parameters. • Some NUR reactor parameters have been analyzed. - Abstract: This paper introduces the Integrated Thermal Hydraulic and Neutronic Analyzer SYStem (ITHNA.SYS) that has been developed for the Algerian research reactor NUR. It is used both as an operating aid tool and as a core physics engineering analysis tool. The system embeds three modules of the MTR-PC software package developed by INVAP SE: the cell calculation code WIMSD, the core calculation code CITVAP and the program TERMIC for thermal hydraulic analysis of a material testing reactor (MTR) core in forced convection. ITHNA.SYS operates both in on-line and off-line modes. In the on-line mode, the system is linked, via the computer parallel port, to the data acquisition console of the reactor control room and allows a real time monitoring of major physical and safety parameters of the NUR core. PC-based ITHNA.SYS provides a viable and convenient way of using an accumulated and often complex reactor physics stock of knowledge and frees the user from the intricacy of adequate reactor core modeling. This guaranties an accurate, though rapid, determination of a variety of neutronic and thermal hydraulic parameters of importance for the operation and safety analysis of the NUR research reactor. Instead of the several hours usually required, the processing time for the determination of such parameters is now reduced to few seconds. Validation of the system was performed with respect to experimental measurements and to calculations using reference codes. ITHNA.SYS can be easily adapted to accommodate other kinds of MTR reactors.

  18. Evaluation of critical heat flux performances for design strategy of new research reactor nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Bang, In Cheol; Lee, Kwi Lim; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2006-02-15

    The present project investigated stable burnout heat flux correlations applicable to research reactor operation conditions of low pressure, low temperature and high flow rate. In addition, in series of thermal limits important to safety of the reactor, ONB and OFI correlations also were investigated. There are some world CHF databases for tube-inside flow. In order to design a research reactor, DNB is final design limit factor and so the collection of the data or correlation are very important. The optimal core cooling capability can be done by considering neutronics, economical efficiency, materials limit together through engineering judgement based on DNB correlations. The project collected the materials and correlations applicable to research reactor conditions. The correlations give a fundamental base for analyzing thermal limit factors and will be used helpfully in review of regulatory body and designer for safety evaluation.

  19. Production and release rate of 37Ar from the UT TRIGA Mark-II research reactor.

    Science.gov (United States)

    Johnson, Christine; Biegalski, Steven R; Artnak, Edward J; Moll, Ethan; Haas, Derek A; Lowrey, Justin D; Aalseth, Craig E; Seifert, Allen; Mace, Emily K; Woods, Vincent T; Humble, Paul

    2017-02-01

    Air samples were taken at various locations around The University of Texas at Austin's TRIGA Mark II research reactor and analyzed to determine the concentrations of 37Ar, 41Ar, and 133Xe present. The measured ratio of 37Ar/41Ar and historical records of 41Ar releases were then utilized to estimate an annual average release rate of 37Ar from the reactor facility. Using the calculated release rate, atmospheric transport modeling was performed in order to determine the potential impact of research reactor operations on nearby treaty verification activities. Results suggest that small research reactors (∼1 MWt) do not release 37Ar in concentrations measurable by currently proposed OSI detection equipment. Copyright © 2016 Elsevier Ltd. All rights reserved.

  20. Reactor safety research programs. Quarterly report, January-March 1982

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S.K. (ed.)

    1982-07-01

    This document summarizes work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1982, for the Division of Accident Evaluation and the Division of Engineering Technology, US Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities.

  1. Reactor safety research programs. Quarterly report, April-June 1982

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S.K. (ed.)

    1982-11-01

    This document summarizes work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1982, for the Division of Accident Evaluation and the Division of Engineering Technology, US Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities.

  2. Event management in research reactors; Gestion de eventos en reactores de investigacion

    Energy Technology Data Exchange (ETDEWEB)

    Perrin, C.D. [Coordinador Reactores de Investigacion y Conjuntos Criticos, Autoridad Regulatoria Nuclear (Argentina)]. e-mail: cperrin@sede.arn.gov.ar

    2006-07-01

    In the Radiological and Nuclear Safety field, the Nuclear Regulatory Authority of Argentina controls the activities of three investigation reactors and three critical groups, by means of evaluations, audits and inspections, in order to assure the execution of the requirements settled down in the Licenses of the facilities, in the regulatory standards and in the documentation of mandatory character in general. In this work one of the key strategies developed by the ARN to promote an appropriate level of radiological and nuclear safety, based on the control of the administration of the abnormal events that its could happen in the facilities is described. The established specific regulatory requirements in this respect and the activities developed in the entities operators are presented. (Author)

  3. Recent results of research on supercritical water-cooled reactors in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Starflinger, J.; Koehly, C.; Schulenberg, T. [Forschungszentrum Karlsruhe, Karlsruhe (Germany); Maraczy, C. [AEKI-KFKI, Budapest (Hungary); Toivonen, A.; Penttila, S. [VTT Technical Research Centre, Espoo (Finland); Chandra, L.; Lycklama a Nijeholt, J.A. [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands)

    2009-07-01

    In Europe, the research on Supercritical Water-Cooled Reactors is integrated in a project called 'High Performance Light Water Reactor Phase 2' (HPLWR Phase 2), co-funded by the European Commission. Ten partners and three active supporters are working on critical scientific issues to determine the potential of this reactor concept in the electricity market. The recent design of the HPLWR including flow paths is described in this paper. Exemplarily, design analyses are presented addressing neutronics, thermal-hydraulics, thermo-mechanics, materials investigations and heat transfer. (author)

  4. Fuel-coolant interaction (FCI) phenomena in reactor safety. Current understanding and future research needs

    Energy Technology Data Exchange (ETDEWEB)

    Speis, T.P. [Maryland Univ., College Park, MD (United States); Basu, S.

    1998-01-01

    This paper gives an account of the current understanding of fuel-coolant interaction (FCI) phenomena in the context of reactor safety. With increased emphasis on accident management and with emerging in-vessel core melt retention strategies for advanced light water reactor (ALWR) designs, recent interest in FCI has broadened to include an evaluation of potential threats to the integrity of reactor vessel lower head and ex-vessel structural support, as well as the role of FCI in debris quenching and coolability. The current understanding of FCI with regard to these issues is discussed, and future research needs to address the issues from a risk perspective are identified. (author)

  5. Estimation of Na-24 activity concentration in BAEC TRIGA Research Reactor

    Directory of Open Access Journals (Sweden)

    M. Ajijul Hoq

    Full Text Available The Bangladesh Atomic Energy Commission (BAEC TRIGA Research Reactor is a unique nuclear installation of the country generally implemented for a wide variety of research applications and serves as an excellent source of neutron. During reactor operation it is necessary to measure and control the activity concentration of the pool water for fuel element failure detection and for the determination of contamination. The present study deals with the estimation of activity concentration for Na-24 present in water coolant produced as a result of 23Na (n, γ 24Na reaction. Several governing equations have been employed to estimate the Na-24 activity concentrations theoretically at different reactor power levels including maximum reactor power of 2.4 MW. From the obtained result it is ensured that the estimated Na-24 activity of 8.83 × 10−3 μCi/cm3 is not significant enough for any radiological hazard. Thus for ensuring radiological safety issues of the research reactor the assessment performed under the present study has an implication. Keywords: TRIGA Research Reactor, Na-24, Activity concentration, Purification system, Water impurity, Radiological safety

  6. MITR-III: Upgrade and relicensing studies for the MIT Research Reactor. Second annual report

    Energy Technology Data Exchange (ETDEWEB)

    Trosman, H.G. [ed.; Lanning, D.D.; Harling, O.K.

    1994-08-01

    The current operating license of the MIT research reactor will expire on May 7, 1996 or possibly a few years later if the US Nuclear Regulatory Commission agrees that the license period can start with the date of initial reactor operation. Driven by the imminent expiration of the operating license, a team of nuclear engineering staff and students have begun a study of the future options for the MIT Research Reactor. These options have included the range from a major rebuilding of the reactor to its decommissioning. This document reports the results of a two year intensive activity which has been supported by a $148,000 grant from the USDOE contract Number DEFG0293ER75859, approximately $100,000 of internal MIT funds and Nuclear Engineering Department graduate student fellowships as well as assistance from international visiting scientists and engineers.

  7. Design Guideline for Primary Heat Exchanger in a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sunil; Seo, Kyoung-Woo; Kim, Seong-Hoon; Chi, Dae-Young; Park, Cheol [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this paper, analytical study is conducted to track the variation of the PCS outlet temperature in conditions of the constant core power and constant SCS inlet temperature. The PCS circulates demineralized water to remove the heat generated in reactor core. The heat is transferred to the cold water of the SCS through the primary heat exchanger. In JRTR, Plate-type Heat Exchanger (PHE) was used as the primary heat exchanger. The cooling tower automatically sets the SCS inlet temperature constant by fan speed control. The flow rate of SCS is adjusted to be identical with the PCS flow rate. To design the PHE, the inlet and outlet temperatures and the flow rates for both systems should be determined. The flow rate has the allowable band for the safe operation from the lower limit to upper limit resulting in different temperature distribution in the PHE. Specially, the PCS outlet temperature which is the core inlet temperature is used for a safety parameter for the reactor shutdown. Therefore, we need to figure out which limit for the flow rate should be used from the conservative point of view. At 200 kg/s of PCS and SCS flow rates, the inlet and outlet temperatures are 41.3℃and 34℃, respectively. With increase of the flow rate, both of PCS inlet and outlet temperatures decrease to 33.6℃ and 39.9℃. This result means the low limit of the allowable flow band should be used for the conservative design of primary heat exchanger. If the upper limit of the allowable flow band is used, the PCS outlet temperature which is the safety parameter used for the reactor shutdown increases with decrease of the flow rate.

  8. Quality management in BNCT at a nuclear research reactor.

    Science.gov (United States)

    Sauerwein, Wolfgang; Moss, Raymond; Stecher-Rasmussen, Finn; Rassow, Jürgen; Wittig, Andrea

    2011-12-01

    Each medical intervention must be performed respecting Health Protection directives, with special attention to Quality Assurance (QA) and Quality Control (QC). This is the basis of safe and reliable treatments. BNCT must apply QA programs as required for performance and safety in (conventional) radiotherapy facilities, including regular testing of performance characteristics (QC). Furthermore, the well-established Quality Management (QM) system of the nuclear reactor used has to be followed. Organization of these complex QM procedures is offered by the international standard ISO 9001:2008. Copyright © 2011 Elsevier Ltd. All rights reserved.

  9. Water treatment process in the JEN-1 Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Urgel, M.; Perez-Bustamante, J. A.; Batuecas, T.

    1965-07-01

    The main characteristics and requirements which must be met with by waters to be used for nuclear reactors were studied paying attention separately both to those used in primary and secondary circuits as well as to the purification systems to be employed in each case. The experiments carried out for the initial pretreatment of water and the ion-exchange de ionization processes including a number of systems consisting of separated and mixed beds loaded with a variety of different commercially available resins are described. (Author) 24 refs.

  10. Operational performance of the three bean salad control algorithm on the ACRR (Annular Core Research Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Ball, R.M.; Madaras, J.J. (B and W Nuclear Technologies, Lynchburg, VA (USA). Space and Defense Systems); Trowbridge, F.R. Jr.; Talley, D.G.; Parma, E.J. Jr. (Sandia National Labs., Albuquerque, NM (USA))

    1991-01-01

    Experimental tests on the Annular Core Research Reactor have confirmed that the Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute. 3 refs., 4 figs., 1 tab.

  11. A pragmatic approach towards designing a second shutdown system for Tehran research reactor

    OpenAIRE

    Boustani Ehsan; Khakshournia Samad; Khalafi Hossein

    2016-01-01

    One second shutdown system is proposed for the Tehran Research Reactor to achieve the goal of higher safety in compliance with current operational requirements and regulations and improve the overall reliability of the reactor shutdown system. The proposed second shutdown system is a diverse, independent shutdown system compared to the existing rod based one that intends to achieve and maintain sub-criticality condition with an enough shutdown margin in man...

  12. Numerical Research on Hybrid Fuel Locking Device for Upward Flow Core-Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hyung; Cho, Yeong-Garp; Yoo, Yeon-Sik; Ryu, Jeong-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The assembly must be held firmly against these forces, but cannot be permanently attached to the support stand because periodic refueling of the reactor requires removal or relocation of each assembly. There are so many kinds of fuel locking device, but they are operated manually. As a part of a new project, we have investigated a hybrid fuel locking device (HFLD) for research reactor which is operated automatically. Prior method of holding down the fuel assembly includes a hybrid zero electromagnet consisting of an electromagnet and a permanent magnet. The role of an electromagnet is converged to zero power for overcoming the lifting power of a permanent magnet by controlling the coil current. At this time, a HFLD is an unlocking state. On the contrary, it is locking state that only a permanent magnet works when the power of an electromagnet is off. The results of a FEM in this work lead to the following conclusions: (1) It is possible that an electromagnet is converged to zero power for overcoming the lifting power of a permanent magnet by remote controlling the coil current. (2) At this time, it is able to detect remotely using proximity sensor whether a HFLD is latched or not.

  13. Air leakage analysis of research reactor HANARO building in typhoon condition for the nuclear emergency preparedness

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Goany Up; Lee, Hae Cho; Kim, Bong Seok; Kim, Jong Soo; Choi, Pyung Kyu [Dept. of Emergency Preparedness, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. It was found that the leak rate is 0.1%·day{sup -1} of air, 0.004%·day{sup -1} of noble gas and 3.7×10{sup -5}%·day{sup -1} of aerosol during typhoon passing. The air leak rate of 0.1%·day can be converted into 1.36 m{sup 3}·hr{sup -1} , but the design leak rate in HANARO safety analysis report was considered as 600 m3·hr{sup -1} under the condition of 20 m·sec{sup -1} wind speed outside of the building by typhoon. Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.

  14. Smoking among in-school adolescents in Dar Es Salam, Tanzania ...

    African Journals Online (AJOL)

    Using data from the 2003 Dar es Salaam (Tanzania) Global Youth Tobacco Survey (GYTS), we assessed factors associated with current cigarette smoking among adolescents. We estimated frequencies and conducted logistic regression analysis to identify predictors of current cigarette smoking. Of the 1947 students ...

  15. Safety Issues at the DOE Test and Research Reactors. A Report to the U.S. Department of Energy.

    Science.gov (United States)

    National Academy of Sciences - National Research Council, Washington, DC. Commission on Physical Sciences, Mathematics, and Resources.

    This report provides an assessment of safety issues at the Department of Energy (DOE) test and research reactors. Part A identifies six safety issues of the reactors. These issues include the safety design philosophy, the conduct of safety reviews, the performance of probabilistic risk assessments, the reliance on reactor operators, the fragmented…

  16. Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately.

  17. Ageing management and refurbishment of Ghana Research Reactor-1 (GHARR-1)

    Energy Technology Data Exchange (ETDEWEB)

    Amponsahabu, Edward Oscar; Gbadago, Joseph Korbla; Addo, Moses Ankamah; Sogbadji, Robert Bright Mawuko; Odoi, Henry Cecil; Gyamfi, Kwame; Ampong, Atta Gyekye; Opate, Nicholas Sackitey [Ghana Atomic Energy Commission, Accra (Ghana)

    2013-07-01

    Ageing management is an essential component of the routine practices at the Ghana Research Reactor-1 (GHARR-1) Facility. The reactor is Miniature Neutron Source Reactor with a rated power of 30 kW. GHARR-1 was installed and attained criticality on December 17, 1994 and commissioned on 8th March, 1995. It has since been in operation. The routine practices and operational procedures have been set out with clear emphasis on ageing policy at the facility. Some electronic components are changed regularly during maintenance sessions and keeping to regular purification of the reactor and pool water to mitigate against corrosion. This paper outlines the ageing management programme, mitigation practices, strategies for ageing management, periodic safety reviews, consideration of ageing during designing, design features for components and unit replacement, top beryllium shim addition, and succession planning. Information sharing with other operating organization is one of the means considered by GHARR-1 to attain excellence.

  18. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  19. Monte Carlo Analysis of the Accelerator-Driven System at Kyoto University Research Reactor Institute

    Directory of Open Access Journals (Sweden)

    Wonkyeong Kim

    2016-04-01

    Full Text Available An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan, a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcroft–Walton type accelerator, which generates the external neutron source by deuterium–tritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.

  20. R and D on fuzzy control applications to the BR1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ruan, D.; Li, X. [Nuclear Research Center, Mol (Belgium)

    1998-12-31

    Fuzzy control applications in nuclear reactor operations present a tremendous challenge. The main reason for this is the public awareness of the risks of nuclear industry and the very strict safety regulations in force for nuclear power plants. The very same regulations prevent a researcher from quickly introducing novel fuzzy-logic methods into this field. On the other hand, the application of fuzzy logic has, despite the ominous sound of the word `fuzzy` to nuclear engineers, a number of very desirable advantages over classical methods, e.g., its robustness and the capability to include human experience into the controller. In this paper we report an on-going R and D project for controlling the power level of the Belgian Nuclear Reactor 1 (BR1) at the Belgian Nuclear Research Centre (SCK.CEN). The project started in 1995 and aims to investigate the added value of fuzzy control for nuclear reactors. We first review some relevant literature on fuzzy logic control in nuclear reactors, then present the state-of-the-art of the BR1 project. After experimenting fuzzy logic control under off-line tests at the BR1 reactor, we now foresee a new development for a closed-loop fuzzy control as an on-line operation of the BR1 reactor. Finally, we present the new development for a closed-loop fuzzy logic control at BR1 with an understanding of the safety requirements for this real fuzzy logic control application in nuclear rectors. (author) 18 refs.

  1. Research and development of an electrochemical biocide reactor

    Science.gov (United States)

    See, G. G.; Bodo, C. A.; Glennon, J. P.

    1975-01-01

    An alternate disinfecting process to chemical agents, heat, or radiation in an aqueous media has been studied. The process is called an electrochemical biocide and employs cyclic, low-level voltages at chemically inert electrodes to pass alternating current through water and, in the process, to destroy microorganisms. The paper describes experimental hardware, methodology, and results with a tracer microorganism (Escherichia coli). The results presented show the effects on microorganism kill of operating parameters, including current density (15 to 55 mA/sq cm (14 to 51 ASF)), waveform of applied electrical signal (square, triangular, sine), frequency of applied electrical signal (0.5 to 1.5 Hz), process water flow rate (100 to 600 cc/min (1.6 to 9.5 gph)), and reactor resident time (0 to 4 min). Comparisons are made between the disinfecting property of the electrochemical biocide and chlorine, bromine, and iodine.

  2. Activity report on the utilization of research reactors. Japanese Fiscal Year, 1998

    Energy Technology Data Exchange (ETDEWEB)

    Toyoda, Masayuki; Koyama, Yoshimi [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-03-01

    This is the second issue of the activity report on the utilization of research reactors in the fields of neutron beam experiments, neutron activation analysis, radioisotope production, etc., performed during Japanese Fiscal Year 1998 (April 1, 1998 - March 31, 1999). All reports in this volume were described by users from JAERI and also users from the other organizations, i.e., universities, national research institutes and private companies, who have utilized our research reactor utilization facilities for the purpose of the above studies. (author)

  3. Activity report on the utilization of research reactors. Japanese Fiscal Year, 1999

    Energy Technology Data Exchange (ETDEWEB)

    Toyoda, Masayuki [ed.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    This is the second issue of the activity report on the utilization of research reactors in the fields of neutron beam experiments, neutron activation analysis, radioisotope production, etc., performed during Japanese Fiscal Year 1999 (April 1, 1999 - March 31, 2000). All reports in this volume were described by users from JAERI and also users from the other organizations, i.e., universities, national research institutes and private companies, who have utilized our research reactor utilization facilities for the purpose of the above studies. (author)

  4. Replacement Nuclear Research Reactor: Draft Environmental Impact Statement. Vol. 1. Main report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-07-01

    The Draft Environmental Impact Statement (EIS) for the replacement of the Australian Research reactor has been released. An important objective of the EIS process is to ensure that all relevant information has been collected and assessed so that the Commonwealth Government can make an informed decision on the proposal. The environmental assessment of the proposal to construct and operate a replacement reactor described in the Draft EIS has shown that the scale of environmental impacts that would occur would be acceptable, provided that the management measures and commitments made by ANSTO are adopted. Furthermore, construction and operation of the proposed replacement reactor would result in a range of benefits in health care, the national interest, scientific achievement and industrial capability. It would also result in a range of benefits derived from increased employment and economic activity. None of the alternatives to the replacement research reactor considered in the Draft EIS can meet all of the objectives of the proposal. The risk from normal operations or accidents has been shown to be well within national and internationally accepted risk parameters. The dose due to reactor operations would continue to be small and within regulatory limits. For the replacement reactor, the principle of `As Low As Reasonably Achievable` would form an integral part of the design and licensing process to ensure that doses to operators are minimized. Costs associated with the proposal are $286 million (in 1997 dollars) for design and construction. The annual operating and maintenance costs are estimated to be $12 million per year, of which a significant proportion will be covered by commercial activities. The costs include management of the spent fuel from the replacement reactor as well as the environmental management costs of waste management, safety and environmental monitoring. Decommissioning costs for the replacement reactor would arise at the end of its lifetime

  5. Prediction of Decommissioning Cost for Kijang Research Reactor Using Power Data of DACCORD

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Yun Jeong; Jin, Hyung Gon; Park, Hee Seong; Park, Seung Kook [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    There are 3 types of cost estimate that can be used, and each have a different level of accuracy: (i) Order of magnitude estimate: One without detailed engineering data, where an estimate is prepared using scale-up or -down factors and approximate ratios. It is likely that the overall scope of the project has not been well defined. The level of accuracy expected is -30% to +50%. The cost plans to predict referring to abroad examples as decommissioning cost estimation has still not developed and been commercial method for Kijang research reactor. In Kijang research reactor case, overall scope of business isn't yet decided. Then it is supposed to estimate cost with type (i). The IAEA project, entitled 'DACCORD' (Data Analysis and Collection for Costing of Research Reactor Decommissioning) performs decommissioning costing after collecting and analyzing the information related to research reactors around the world for several years. Also decommissioning costing method development tends to increase in the each country. This paper aims to estimate preliminary decommissioning cost based on total decommissioning cost per thermal power rate of research reactor presented in DACCORD project' data which is collected by member state. In this paper, preliminary decommissioning cost is estimated based on total decommissioning cost per thermal power rate of research reactor presented in DACCORD data which is collected by member state. Although there exists a general tendency for costs to increase with increasing thermal power, the limited data available show that decommissioning costs at any given power level can vary widely, with increased variability at higher power levels. Variations in decommissioning cost for the research reactors of the same or similar thermal power are caused by differences in reactor types and design, decommissioning project scopes, country- specific unit workforce costs, and other reactor or project factors. An important factor for the

  6. A Development of Technical Specification of a Research Reactor with Plate Fuels Cooled by Upward Flow

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sujin; Kim, Jeongeun; Kim, Hyeonil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The contents of the TS(Technical Specifications) are definitions, safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls. TS for Nuclear Power Plants (NPPs) have been developed since many years until now. On the other hands, there are no applicable modernized references of TS for research reactors with many differences from NPPs in purpose and characteristics. Fuel temperature and Departure from Nuclear Boiling Ratio (DNBR) are being used as references from the thermal-hydraulic analysis point of view for determining whether the design of research reactors satisfies acceptance criteria for the nuclear safety or not. Especially for research reactors using plate-type fuels, fuel temperature and critical heat flux, however, are very difficult to measure during the reactor operation. This paper described the outline of main contents of a TS for open-pool research reactor with plate-type fuels using core cooling through passive systems, where acceptance criteria for nuclear safety such as CHF and fuel temperature cannot be directly measured, different from circumstances in NPPs. Thus, three independent variables instead of non-measurable acceptance criteria: fuel temperature and CHF are considered as safety limits, i.e., power, flow, and flow temperature.

  7. Annual report of department of research reactors, 2001. April 1, 2001 - March 31, 2002

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-12-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2001 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. (author)

  8. Proceedings of the 1998 workshop on the utilization of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-10-01

    The 1998 Workshop on the Utilization of Research Reactors, which is the seventh Workshop on the theme of research reactor utilization was held in Yogyakarta and Serpong, Indonesia from February 8 to 14. This Workshop was executed based on the agreement in the Ninth International Conference for Nuclear Cooperation in Asia (ICNCA) held in Tokyo, March 1998. The whole Workshop consists of the Workshop on the theme of following three fields, 1) Neutron Scattering, 2) Neutron Activation analysis and 3) Safe Operation and Maintenance of Research Reactor, and the Sub-workshop carried out the experiment of Neutron Activation analysis. The total number of participants for the workshop was about 100 people from 8 countries, i.e. Australia, China, Indonesia, Korea, Malaysia, Thailand, Vietnam and Japan. The 38 papers are indexed individually. (J.P.N.)

  9. Annual report of department of research reactor, 1995 (April 1, 1995 - March 31, 1996)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1995 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

  10. Analysis on safeguard approach of radioactive waste at KIJANG research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun Jo; Lee, Sung Ho; Lee, Byung Doo; Kim, In Chul; Kim, Hyun Sook; Jung, Juang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    KIJANG Research Reactor (KJRR) will be constructed in Busan in order to provide the self-sufficiency of RI demand including Mo-99, to increase the neutron transmutation doping (NTD) capacity and to develop and validate technologies related to the research reactor. Considering the categorization of nuclear facility such as item counting and bulk facility, HANARO which is another research reactors in Korea is item counting facility because physical/chemical forms of nuclear material are not changes. During the dissolving process, radioactive wastes containing nuclear material are occurred at KJRR. In this paper, the features of the KJRR are described and safeguards approach on the radioactive wastes containing nuclear material occurred at KJRR are reviewed. This paper reviews the safeguards approach on radioactive wastes containing nuclear materials occurred during FM production at KJRR. Most uranium dissolved during FM production process are collected in U filter cakes and very tiny amount of uranium will be remained in the ILLW.

  11. Annual report of Department of Research Reactor, 1996. April 1, 1996 - March 31, 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1996 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

  12. Annual report of department of research reactor, 1999. April 1, 1999 - March 31, 2000

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization of the JRR-3M (new JRR-3) and the JRR-4 and for the related R and D. Besides the decommissioning of the JRR-2 and RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 1999 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. (author)

  13. Annual report of department of research reactor, 1994. April 1, 1994 - March 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-10-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1994 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as well as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author).

  14. Development of a research reactor power measurement system using Cherenkov radiation

    Energy Technology Data Exchange (ETDEWEB)

    Salles, Brício M.; Mesquita, Amir Z., E-mail: briciomares@hotmail.com, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-11-01

    Nuclear research reactors are usually located in open pools, to allow visibility to the core and bluish luminosity of Cherenkov radiation. Usually the thermal power released in these reactors is monitored by chambers that measure the neutron flux, as it is proportional to the power. There are other methods used for power measurement, such as monitoring the core temperature and the energy balance in the heat exchanger. The brightness of Cherenkov's radiation is caused by the emission of visible electromagnetic radiation (in the blue band) by charged particles that pass through an insulating medium (water in nuclear research reactors) at a speed higher than that of light in this medium. This effect was characterized by Pavel Cherenkov, which earned him the Nobel Prize for Physics in 1958. The project's objective is to develop an innovative and alternative method for monitoring the power of nuclear research reactors. It will be performed by analyzing and monitoring the intensity of luminosity generated by Cherenkov radiation in the reactor core. This method will be valid for powers up to 250 kW, since above that value the luminosity saturates, as determined by previous studies. The reactor that will be used to test the method is the TRIGA, located at Nuclear Technology Development Center (CDTN), which currently has a maximum operating power of 250 kW. This project complies with International Atomic Energy Agency (IAEA) recommendations on reactor safety. It will give more redundancy and diversification in this measure and will not interfere with its operation. (author)

  15. The Need for Cyber-Informed Engineering Expertise for Nuclear Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Robert Stephen [Idaho National Laboratory

    2015-12-01

    Engineering disciplines may not currently understand or fully embrace cyber security aspects as they apply towards analysis, design, operation, and maintenance of nuclear research reactors. Research reactors include a wide range of diverse co-located facilities and designs necessary to meet specific operational research objectives. Because of the nature of research reactors (reduced thermal energy and fission product inventory), hazards and risks may not have received the same scrutiny as normally associated with power reactors. Similarly, security may not have been emphasized either. However, the lack of sound cybersecurity defenses may lead to both safety and security impacts. Risk management methodologies may not contain the foundational assumptions required to address the intelligent adversary’s capabilities in malevolent cyber attacks. Although most research reactors are old and may not have the same digital footprint as newer facilities, any digital instrument and control function must be considered as a potential attack platform that can lead to sabotage or theft of nuclear material, especially for some research reactors that store highly enriched uranium. This paper will provide a discussion about the need for cyber-informed engineering practices that include the entire engineering lifecycle. Cyber-informed engineering as referenced in this paper is the inclusion of cybersecurity aspects into the engineering process. A discussion will consider several attributes of this process evaluating the long-term goal of developing additional cyber safety basis analysis and trust principles. With a culture of free information sharing exchanges, and potentially a lack of security expertise, new risk analysis and design methodologies need to be developed to address this rapidly evolving (cyber) threatscape.

  16. European research activities within the project: High Performance Light Water Reactor phase 2 (HPLWR phase 2)

    Energy Technology Data Exchange (ETDEWEB)

    Starflinger, J.; Schulenberg, T. [Forschungszentrum Karlsruhe GmbH, Institute for Nuclear and Energy Technologies, Karlsruhe (Germany); Marsault, P. [CEA Cadarache (DER/SESI), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Reacteurs; Bittermann, D. [AREVA NP, NEPR-G, Erlangen (Germany); Maraczy, C. [AEKI-KFKI, Budapest (Hungary); Laurien, E. [Stuttgart Univ. IKE (Germany); Lycklama, J.A. [NRG Petten, NL (Netherlands); Anglart, H. [KTH Energy Technology, Stockholm (Sweden); Aksan, N. [Paul Scherrer Institut CH, Villigen PSI (Switzerland); Ruzickova, M. [UJV Rez plc, Husinec-Rez c.p. (Czech Republic); Heikinheimo, L. [VTT, FIN (Finland)

    2007-07-01

    The High Performance Light Water Reactor (HPLWR) is a Light Water Reactor (LWR) operating at supercritical pressure (25 MPa). It belongs to the six reactors currently being investigated under the framework of the Generation IV International Forum. The most visible advantage of the HPLWR shall be the low construction costs in the order of 1000 Euro/kWe, because of size reduction of components and buildings compared to current Light Water Reactors, and the low electricity production costs which are targeted at 3-4 cents/kWh. In Europe, investigations on the HPLWR have been integrated into a joint research project, called High Performance Light Water Reactor Phase 2 (HPLWR Phase 2), which is co-funded by the European Commission. Within 42 months, ten partners from eight European countries working on critical scientific issues shall show the feasibility of the HPLWR concept. This paper reports on 5 points relevant for HPLWR: 1) design and integration, 2) core design, 3) safety, 4) materials, and 5) heat transfer. The final goal is to assess the future potential of this reactor in the electricity market.

  17. Radioisotope radiotherapy research and achievements at the University of Missouri Research Reactor

    Science.gov (United States)

    Ehrhardt, G. J.; Ketring, A. R.; Cutler, C. S.

    2003-01-01

    The University of Missouri Research Reactor (MURR) in collaboration with faculty in other departments at the University of Missouri has been involved in developing new means of internal radioisotopic therapy for cancer for many years. These efforts have centered on methods of targeting radioisotopes such as brachytherapy, embolisation of liver tumors with radioactive microspheres, small-molecule-labelled chelates for the treatment of bone cancer, and various means of radioimmunotherapy or labelled receptor agent targeting. This work has produced two radioactive agents, Sm-153 Quadramet™ and Y-90 TheraSphere™, which have U.S. Food and Drug Administration approval for the palliation of bone cancer pain and treatment of inoperable liver cancer, respectively. MURR has also pioneered development of other beta-emitting isotopes for internal radiotherapy such as Re-186, Re-188, Rh-105, Ho-166, Lu-177, and Pm-149, many of which are in research and clinical trials throughout the U.S. and the world. This important work has been made possible by the very high neutron flux available at MURR combined with MURR's outstanding reliability of operation and flexibility in meeting the needs of researchers and the radiopharmaceutical industry.

  18. Estimation of Na-24 activity concentration in BAEC TRIGA Research Reactor

    Science.gov (United States)

    Ajijul Hoq, M.; Malek Soner, M. A.; Salam, M. A.; Khanom, Salma; Fahad, S. M.

    The Bangladesh Atomic Energy Commission (BAEC) TRIGA Research Reactor is a unique nuclear installation of the country generally implemented for a wide variety of research applications and serves as an excellent source of neutron. During reactor operation it is necessary to measure and control the activity concentration of the pool water for fuel element failure detection and for the determination of contamination. The present study deals with the estimation of activity concentration for Na-24 present in water coolant produced as a result of 23Na (n, γ) 24Na reaction. Several governing equations have been employed to estimate the Na-24 activity concentrations theoretically at different reactor power levels including maximum reactor power of 2.4 MW. From the obtained result it is ensured that the estimated Na-24 activity of 8.83 × 10-3 μCi /cm3 is not significant enough for any radiological hazard. Thus for ensuring radiological safety issues of the research reactor the assessment performed under the present study has an implication.

  19. Improvement of the reactivity computer for HANARO research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Jin; Park, S. J.; Jung, H. S.; Choi, Y. S.; Lee, K. H.; Seo, S. G

    2001-04-01

    The reactivity computer in HANARO has a dedicated neutron detection system for experiments and measurements of the reactor characteristics. This system consists of a personal computer and a multi-function I/O board, and collects the signals from the various neutron detectors. The existing hardware and software are developed under the DOS environment so that they are very inconvenient to use and have difficulties in finding replacement parts. Since the continuity of the signal is often lost when we process the wide rang signal, the need for its improvement has been an issue. The purpose of this project is to upgrade the hardware and software for data collection and processing in order for them to be compatible with Windows{sup TM} operating system and to solve the known issue. We have replaced the existing system with new multi-function I/O board and Pentium III class PC, and the application program for the wide range reactivity measurement and multi-function signal counter have been developed. The newly replaced multi-function I/O board has seven times fast A/D conversion rate and collects sufficient amount of data in a short time. The new application program is user-friendly and provides various useful information on its display screen so that the ability of data processing and storage has been very much enhanced.

  20. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

  1. A reload and startup plan for conversion of the NIST research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-03-31

    The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reload portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.

  2. A Reload and Startup Plan for and #8233;Conversion of the NIST Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Varuttamaseni, A. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2017-09-30

    The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts.The reload portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.

  3. Determination of the optimal positions for installing gamma ray detection systems at Tehran Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sayyah, A. [Department of Radiation Application, Shahid Beheshti University (Iran, Islamic Republic of); Rahmani, F., E-mail: FRahamni@kntu.ac.in [K.N. Toosi University of Technology, Department of Physics (Iran, Islamic Republic of); Khalafi, H. [Nuclear Science and Technology Research Institute (NSTRI) (Iran, Islamic Republic of)

    2015-09-01

    Dosimetric instruments must constantly monitor radiation dose levels in different areas of nuclear reactor. Tehran Research Reactor (TRR) has seven beam tubes for different research purposes. All the beam tubes extend from the reactor core to Beam Port Floor (BPF) of the reactor facility. During the reactor operation, the gamma rays exiting from each beam tube outlet produce a specific gamma dose rate field in the space of the BPF. To effectively monitor the gamma dose rates on the BPF, gamma ray detection systems must be installed in optimal positions. The selection of optimal positions is a compromise between two requirements. First, the installation positions must possess largest gamma dose rates and second, gamma ray detectors must not be saturated in these positions. In this study, calculations and experimental measurements have been carried out to identify the optimal positions of the gamma ray detection systems. Eight three dimensional models of the reactor core and related facilities corresponding to eight scenarios have been simulated using MCNPX Monte Carlo code to calculate the gamma dose equivalent rate field in the space of the BPF. These facilities are beam tubes, thermal column, pool, BPF space filled with air, facilities such as neutron radiography facility, neutron powder diffraction facility embedded in the beam tubes as well as biological shields inserted into the unused beam tubes. According to the analysis results of the combined gamma dose rate field, three positions on the north side and two positions on the south side of the BPF have been recognized as optimal positions for installing the gamma ray detection systems. To ensure the consistency of the simulation data, experimental measurements were conducted using TLDs (600 and 700) pairs during the reactor operation at 4.5 MW.

  4. Development of a Monolithic Research Reactor Fuel Type at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Clark, C.R.; Briggs, R.J.

    2004-10-06

    The Reduced Enrichment for Research and Test Reactors (RERTR) program has been tasked with the conversion of research reactors from highly enriched to low-enriched uranium (LEU). To convert several high power reactors, monolithic fuel, a new fuel type, is being developed. This fuel type replaces the standard fuel dispersion with a fuel alloy foil, which allows for fuel densities far in excess of that found in dispersion fuel. The single-piece fuel foil also contains a significantly lower interface area between the fuel and the aluminum in the plate than the standard fuel type, limiting the amount of detrimental fuel-aluminum interaction that can occur. Implementation of monolithic fuel is dependant on the development of a suitable fabrication method as traditional roll-bonding techniques are inadequate.

  5. Decontamination and decommissioning project of the TRIGA Mark-2 and 3 research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jung, K. J.; Baik, S. T.; Chung, U. S.; Jung, K. H.; Park, S. K.; Lee, B. J.; Kim, J. K.; Yang, S. H

    2000-01-01

    During the review on the decommissioning plan and environmental impact assessment report by the KINS, the number of the inquired items were two hundred and fifty one, and the answers were made and sent until September 10, 1999, as the screened review results were reported to Ministry of Science and Technology(MOST) in December 14, 1999, all the reviews on the licence were over. Radioactive liquid wastes of 400 tons generated during the operation of the research reactors including reactor vessels are stored in the facility of the research reactor 1 and 2. Those liquid wastes have the low-level-radioactivity which can be discharged to the surroundings, but was wholly treated to be vaporized naturally by means of the increased numbers of the natural vaporization disposal facilities with the annual capacity of 200 tons for the purpose of the minimized environmental contamination.

  6. Corrosion of spent fuels from research and prototype reactors under conditions relevant to geological disposal

    Energy Technology Data Exchange (ETDEWEB)

    Curtius, Hilde; Bosbach, Dirk; Deissmann, Guido [Forschungszentrum Juelich GmbH (Germany). Inst. for Nuclear Waste Management and Reactor Safety (IEK-6)

    2015-07-01

    The reference inventory of high-level nuclear wastes designated for geological disposal in Germany as used within the preliminary safety assessment for a geological repository in the Gorleben salt dome (''vorlaeufige Sicherheitsanalyse Gorleben'', vSG) includes various types of spent nuclear fuels from research and prototype reactors, besides LWR spent fuels and vitrified high-level wastes. This paper will discuss the results of and conclusions from corrosion experiments on spent fuels from prototype high-temperature reactors (HTR) and research reactors that were performed under conditions relevant for a deep geological repository and provided the basis for the derivation of respective source terms in the vSG.

  7. Proceedings of the 1999 workshop on the utilization of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-01

    The 1999 workshop on the utilization of reactors, which is the eighth workshop on the theme of research reactor utilization was held at JAERI Tokai and Mito Plaza Hotel, in Japan from November 25 to December 2. This workshop was executed based on the agreement in the Tenth International conference for Nuclear Cooperation in Asia (ICNCA) held in Tokyo, March 1999. The whole workshop consists of the workshop on the theme of following three fields, 1) neutron scattering, 2) radioisotope production and 3) safe operation and maintenance of research reactor, and the sub-workshop carried out the experiments of small angle neutron scattering. The total number of participants for the workshop was about 70 people from 9 countries, i.e. Australia, China, Indonesia, Korea, Malaysia, The Philippines, Thailand, Vietnam and Japan. The 37 of the presented papers are indexed individually. (J.P.N.)

  8. Initial verification and validation of RAZORBACK - A research reactor transient analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Talley, Darren G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    This report describes the work and results of the initial verification and validation (V&V) of the beta release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This initial V&V effort was intended to confirm that the code work to-date shows good agreement between simulation and actual ACRR operations, indicating that the subsequent V&V effort for the official release of the code will be successful.

  9. Development of a Low Temperature Irradiation Capsule for Research Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Kee Nam; Cho, Man Soon; Lee, Cheol Yong; Yang, Sung Woo; Shin, Yoon Taek; Park, Seng Jae; Kang, Suk Hoon; Kang, Young Hwan; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    A new capsule design was prepared and tested at HANARO for a neutron irradiation of core materials of research reactors as a part of the research reactor development project. Irradiation testing of the materials including graphite, beryllium, and zircaloy-4 that are supposed to be used as core materials in research reactors was required for irradiation at up to 8 reactor operation cycles at low temperature (<100 .deg. C). Therefore, three instrumented capsules were designed and fabricated for an evaluation of the neutron irradiation properties of the core materials (Graphite, Be, Zircaloy-4) of research reactors. The capsules were first designed and fabricated to irradiate materials at low temperature (<100 .deg. C) for a long cycle of 8 irradiation cycles at HANARO. Therefore, the safety of the new designed capsule should be fully checked before irradiation testing. Out-pile performance and endurance testing before HANARO irradiation testing was performed using a capsule under a 110% condition of a reactor coolant flow amount. The structural integrity of the capsule was analyzed in terms of a vibration-induced fatigue cracking of a rod tip of the capsule that is suspected to be the most vulnerable part of a capsule. Another two capsules were irradiated at HANARO for 4 cycles, and one capsule was transferred to a hot cell to examine the integrity of the rod tip of the capsule. After confirming the soundness of the 4 cycle-irradiated capsule, the remaining capsule was irradiated at up to 8 cycles at HANARO. Based on the structural integrity analysis of the capsule, an improved capsule design will be suggested for a longer irradiation test at HANARO.

  10. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  11. Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site.

  12. Experimental computer-controlled instrumentation system for the research reactor DR2

    DEFF Research Database (Denmark)

    Goodstein, L.P.

    1969-01-01

    An instrumentation system has been developed for one of the Danish Atomic Energy Commission's research reactors as part of an experiment on the advantages to be gained by the use of digital computers in a process plant application. Problem areas to be investigated include (a) reliability and safety...

  13. Annual report of department of research reactor, 2000. April 1, 2000 - March 31, 2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-02-01

    The Department of Research Reactor is responsible for the operation, Maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2000 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI Production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. Although the term 'JRR-3M' was used to denote the JRR-3M modified 1990 until the 2000 annual report of the Department of Research Reactor, the term 'JRR-3' will be used from this annual report because the JRR-3 has been operated for about 10 years since the modification and is now under further modification and upgrading study. (author)

  14. 75 FR 62892 - Massachusetts Institute of Technology Research Reactor Environmental Assessment and Finding of No...

    Science.gov (United States)

    2010-10-13

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Massachusetts Institute of Technology Research Reactor Environmental Assessment and Finding of No Significant Impact Correction In notice document 2010-24809 beginning on page 61220 in the issue of Monday...

  15. Status of DOE efforts to renew acceptance of foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Head, C.R.

    1997-08-01

    This presentation summarizes the efforts being made by the Department of Energy to renew acceptance of spent nuclear fuel shipments from foreign research reactors. The author reviews the actions undertaken in this process in a fairly chronological manner, through the present time, as well as the development of an environmental impact statement to support the proposed actions.

  16. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  17. Irradiation of Electronic Components and Circuits at the Portuguese Research Reactor: Lessons Learned

    Science.gov (United States)

    Marques, J. G.; Ramos, A. R.; Fernandes, A. C.; Santos, J. P.

    2016-06-01

    A program was started in 1999 in the Portuguese Research Reactor to test electronics components and circuits for the LHC facility at CERN, initially with a dedicated in-pool irradiation container used at a reactor power of 2 kW and later an irradiation chamber outside the pool, with tailored neutron and gamma filters that could be used at 1 MW. Practice has shown the need to introduce several improvements to the irradiation procedures and infrastructures over the years. In this paper, we review the lessons learned and the major improvements introduced.

  18. Analysis of the Jamaican Slowpoke-2 Research Reactor for the Conversion from HEU to LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Puig, F.; Dennis, Haile T.

    2014-01-01

    The Jamaican SLOWPOKE-2 (JM-1) is a 20 kW research reactor manufactured by Atomic Energy of Canada Limited that has been operating for 30 years at the University of the West Indies, Mona Campus in Kingston, Jamaica. The University, with IAEA assistance under the GTRI/RERTR program, is currently in the process of converting from HEU to LEU. Full-reactor neutronic and thermal hydraulic analyses were performed, using MCNP5 and PLTEMP/ANL v4.1 respectively, on both the existing HEU and proposed LEU core configurations. Although conversion will result in the full nominal reactor power increasing from 20 kW to approximately 22 kW, in order to maintain the 1012 n·cm-2 s-1 flux in the inner irradiation channels, and maximum fuel temperature to increase from ~82°C to ~113°C, the analysis illustrates that increased safety margins will be obtained. No significant reactor behavior changes are expected and the characteristic SLOWPOKE-2 reactor inherent safety features will be preserved.

  19. Experimental estimation of moderator temperature coefficient of reactivity of the IPEN/MB-01 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Rubens C. da; Bitelli, Ulysses D.; Mura, Luiz Ernesto C., E-mail: rubensrcs@usp.br, E-mail: ubitelli@ipen.br, E-mail: credidiomura@gmail.com [Universidade de Sao Paulo (PNV/POLI/USP), SP (Brazil). Arquitetura Naval e Departamento de Engenharia Oceanica; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28 x 26 rectangular array of UO{sub 2} fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D{sub 2}O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4 °C, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and α{sup M}{sub T}(T)(Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them. (author)

  20. Determination of the Design Speed of the Primary Cooling Pump in the Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Hyungi; Seo, Kyoungwoo; Chi, Daeyoung; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    An open-pool type research reactor is widely designed in consideration of the reactor operation and accessibility. Reactor structure assembly is generally placed at the pool bottom. rimary cooling system circulates the coolant from the reactor core to the heat exchanger. Therefore the heat generated from the reactor core is continuously removed. After the primary cooling pumps stop, the decay heat is removed by the coastdown flow induced by the inertia force of a flywheel attached to each primary cooling pump. A pump coastdown flow means that the pump operates with the angular momentums of the shaft, impeller, and flywheel when a loss of electricity occurs. The primary cooling pump consists of the pump, flywheel, and moto. They are connected by flexible couplings. The primary cooling pump is conceptually designed based on the required flow rate and system constraints. A centrifugal pump of Case 1 with a non-dimensional specific speed of 0.59 and specific diameter of 4.94 is chosen as the primary cooling pump based on the hydraulic performance and mechanical integrity.

  1. Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun

    2005-07-15

    Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.

  2. Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor, Rev. 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Hoon; Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun

    2005-12-15

    Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.

  3. 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.

  4. Risk of the research reactor BER II in Berlin; Risiken des Berliner Experimentierreaktors BER II

    Energy Technology Data Exchange (ETDEWEB)

    Paulitz, Henrik; Hoevener, Barbara; Rosen, Alex

    2015-04-20

    The research reactor BER II is sited at the periphery of Berlin in the neighborhood of residential areas. The operational license is limited until December 31, 2019. The reactor is funded by the Federal Government (90%) and the city of Berlin (10%). The stress test has shown that the reactor is not secured against an aircraft crash (airliner or fast flying military jet), meltdown with remarkable radiological consequences to the public would be the consequence. Further hazards result from the radioactive waste transport, explosions and fires. The emergency measures cannot be considered to be sufficient. The city of Berlin would not be able to fulfill the required measures in case of a radiation accident.

  5. Anti-neutrino flux in a research reactor for non-proliferation application

    Energy Technology Data Exchange (ETDEWEB)

    Khakshournia, Samad; Foroughi, Shokoufeh [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Atomic Energy Organization of Iran (AEOI)

    2017-11-15

    Owing to growing interest in the study of emitted antineutrinos from nuclear reactors to test the Atomic Energy Agency safeguards, antineutrino flux was studied in the Tehran Research Reactor (TRR) using ORIGEN code. According to our prediction, antineutrino rate was obtained 2.6 x 10{sup 17} (v{sub e}/sec) in the core No. 57F of the TRR. Calculations indicated that evolution of antineutrino flux was very slow with time and the performed refueling had not an observable effect on antineutrino flux curve for a 5 MW reactor with the conventional refueling program. It is seen that for non-proliferation applications the measurement of the contribution of {sup 239}Pu to the fission using an antineutrino detector is not viable in the TRR.

  6. Radiological survey support activities for the decommissioning of the Ames Laboratory Research Reactor Facility, Ames, Iowa

    Energy Technology Data Exchange (ETDEWEB)

    Wynveen, R.A.; Smith, W.H.; Sholeen, C.M.; Justus, A.L.; Flynn, K.F.

    1984-09-01

    At the request of the Engineering Support Division of the US Department of Energy-Chicago Operations Office and in accordance with the programmatic overview/certification responsibilities of the Department of Energy Environmental and Safety Engineering Division, the Argonne National Laboratory Radiological Survey Group conducted a series of radiological measurements and tests at the Ames Laboratory Research Reactor located in Ames, Iowa. These measurements and tests were conducted during 1980 and 1981 while the reactor building was being decontaminated and decommissioned for the purpose of returning the building to general use. The results of these evaluations are included in this report. Although the surface contamination within the reactor building could presumably be reduced to negligible levels, the potential for airborne contamination from tritiated water vapor remains. This vapor emmanates from contamination within the concrete of the building and should be monitored until such time as it is reduced to background levels. 2 references, 8 figures, 6 tables.

  7. Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

    2001-04-01

    The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

  8. Graphic-object information system {open_quotes}research base for reactor materials science{close_quotes}

    Energy Technology Data Exchange (ETDEWEB)

    Markina, N.V.; Lebedeva, E.E.; Arkhangel`skii, N.V.; Semenov, S.B.; Moiseev, A.L.

    1994-11-01

    An information system developed for reactor materials research is described. The information system incorporates an expert system, MATREKS, and a heirarchial data base. The data base contains information from 20 Russian research reactors. The information system structure, data base structure, search methods, system output modes, and technical facilities and software required are briefly discussed. 6 refs., 2 figs.

  9. 77 FR 13376 - Notice of License Termination for the University of Arizona Research Reactor, License No. R-52

    Science.gov (United States)

    2012-03-06

    ... COMMISSION Notice of License Termination for the University of Arizona Research Reactor, License No. R-52 The... No. R-52, for the University of Arizona Research Reactor (UARR). The NRC has terminated the license... released for unrestricted use. Therefore, Facility Operating License No. R-52 is terminated. For further...

  10. Major update of Safety Analysis Report for Thai Research Reactor-1/Modification 1

    Energy Technology Data Exchange (ETDEWEB)

    Tippayakul, Chanatip [Thailand Institute of Nuclear Technology, Bangkok (Thailand)

    2013-07-01

    Thai Research Reactor-1/Modification 1 (TRR-1/M1) was converted from a Material Testing Reactor in 1975 and it had been operated by Office of Atom for Peace (OAP) since 1977 until 2007. During the period, Office of Atom for Peace had two duties for the reactor, that is, to operate and to regulate the reactor. However, in 2007, there was governmental office reformation which resulted in the separation of the reactor operating organization from the regulatory body in order to comply with international standard. The new organization is called Thailand Institute of Nuclear Technology (TINT) which has the mission to promote peaceful utilization of nuclear technology while OAP remains essentially the regulatory body. After the separation, a new ministerial regulation was enforced reflecting a new licensing scheme in which TINT has to apply for a license to operate the reactor. The safety analysis report (SAR) shall be submitted as part of the license application. The ministerial regulation stipulates the outlines of the SAR almost equivalent to IAEA standard 35-G1. Comparing to the IAEA 35-G1 standard, there were several incomplete and missing chapters in the original SAR of TRR1/M1. The major update of the SAR was therefore conducted and took approximately one year. The update work included detail safety evaluation of core configuration which used two fuel element types, the classification of systems, structures and components (SSC), the compilation of detail descriptions of all SSCs and the review and evaluation of radiation protection program, emergency plan and emergency procedure. Additionally, the code of conduct and operating limits and conditions were revised and finalized in this work. A lot of new information was added to the SAR as well, for example, the description of commissioning program, information on environmental impact assessment, decommissioning program, quality assurance program and etc. Due to the complexity of this work, extensive knowledge was

  11. RETU. The Finnish research programme on reactor safety. Interim report 1995 - May 1997

    Energy Technology Data Exchange (ETDEWEB)

    Vanttola, T.; Puska, E.K. [VTT Energy, Espoo (Finland). Nuclear Energy] [eds.

    1997-08-01

    The Finnish national research programme on Reactor Safety (RETU, 1995-1998) concentrates on the search of safe limits of nuclear fuel and the reactor core, accident management methods and risk management of the operation of nuclear power plants. The annual volume of the programme has been about 26 person years and the annual funding FIM 15 million. This report summarises the structure and objectives of the programme, research fields included and the main results obtained during the period 1995 - May 1997. In the field of operational margins of a nuclear reactor, the behaviour of high burnup nuclear fuel is studied both in normal operation and during power transients. The static and dynamic reactor analysis codes are developed and validated to cope with new fuel designs and complicated three-dimensional reactivity transients and accidents. Research on accident management aims at development and validation of calculation methods needed to plan preventive measures and to train the personnel to severe accident mitigation. Other goals are to reduce uncertainties in phenomena important in severe accidents and to study actions planned for accident management. In the field of risk management probabilistic methods are developed for safety related decision making and for complex phenomena and event sequences. Effects of maintenance on nuclear power plant safety are studied and more effective methods for the assessment of human reliability and safety critical organisations are searched. 135 refs.

  12. Studies on capacity management for factories of nuclear fuel for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Negro, Miguel Luiz Miotto; Durazzo, Michelangelo; Mesquita, Marco Aurélio de; Carvalho, Elita Fontenele Urano de; Andrade, Delvonei Alves de, E-mail: mlnegro@ipen.br, E-mail: mdurazzo@ipen.br, E-mail: elitaucf@ipen.br, E-mail: delvonei@ipen.br, E-mail: mamesqui@usp.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), SP (Brazil). Escola Politécnica. Departamento de Engenharia de Produção

    2017-11-01

    The use and the power of nuclear reactors for research and materials testing is increasing worldwide. That implies the demand for nuclear fuel for this kind of reactors is rising. Thus, the production facilities of this kind of fuel need reliable guidance on how to augment their production in order to meet the increasing demand efficiently, safely and keeping good quality. Focus is given to factories that produce plate type fuel elements loaded with LEU U{sub 3}Si{sub 2}-Al fuel, which are typically used in nuclear research reactors. Of the various production routes for this kind of fuel, we chose the route which uses hydrolysis of uranium hexafluoride. Raising the capacity of this kind of plants faces several problems, especially regarding safety against nuclear criticality. Some of these problems are briefly addressed. The new issue of the paper is the application of knowledge from the area of production administration to the fabrication of nuclear fuel for research reactors. A specific method for the increase in production capacity is proposed. That method was tested by means of discrete event simulation. The data were collected from the nuclear fuel factory at IPEN. The results indicated the proposed method achieved its goal as well as ways of raising production capacity in up to 50%. (author). (author)

  13. A pragmatic approach towards designing a second shutdown system for Tehran research reactor

    Directory of Open Access Journals (Sweden)

    Boustani Ehsan

    2016-01-01

    Full Text Available One second shutdown system is proposed for the Tehran Research Reactor to achieve the goal of higher safety in compliance with current operational requirements and regulations and improve the overall reliability of the reactor shutdown system. The proposed second shutdown system is a diverse, independent shutdown system compared to the existing rod based one that intends to achieve and maintain sub-criticality condition with an enough shutdown margin in many of abnormal situations. It is designed as much as practical based on neutron absorber solution injection into the existing core while the changes and interferences with the existing core structure are kept to a minimum. Core neutronic calculations were performed using MCNPX 2.6.0 and MTR_PC package for the current operational core equipped with the second shutdown system, and one experiment was conducted in the Tehran Research Reactor to test the neutronic calculations. A good agreement was seen between theoretical results and experimental ones. In addition, capability of the second shutdown system in the case of occurrence of design basis accident in the Tehran Research Reactor is demonstrated using PARET program.

  14. Comparison of AFRRI (Armed Forces Radiobiology Research Institute) and ETCA (Etablissement Technique Central de l’Armament) Dosimetry Measurements at AFRRI TRIGA (Training Research Isotopes General Atomic) Reactor

    Science.gov (United States)

    1987-11-01

    TECHNICAL REPORT Comparison of AFRRI and ETCA dosimetry measurements at AFRRI TRIGA reactor M. Dooley G. H. Zeman DEFENSE NUCLEAR AGENCY ARMED...Research Institute (AFRRI) TRIGA (Training Research Isotopes General Atomic) reactor on 10-12 September 1985. The purpose of the experiment was to...anthropomorphic dosimetry phantom (a Incite cylinder 30 cm in diameter and 60 cm tall) in exposure room 1 of the AFRRI TRIGA reactor. As reference

  15. Application of best estimate plus uncertainty in review of research reactor safety analysis

    Directory of Open Access Journals (Sweden)

    Adu Simon

    2015-01-01

    Full Text Available To construct and operate a nuclear research reactor, the licensee is required to obtain the authorization from the regulatory body. One of the tasks of the regulatory authority is to verify that the safety analysis fulfils safety requirements. Historically, the compliance with safety requirements was assessed using a deterministic approach and conservative assumptions. This provides sufficient safety margins with respect to the licensing limits on boundary and operational conditions. Conservative assumptions were introduced into safety analysis to account for the uncertainty associated with lack of knowledge. With the introduction of best estimate computational tools, safety analyses are usually carried out using the best estimate approach. Results of such analyses can be accepted by the regulatory authority only if appropriate uncertainty evaluation is carried out. Best estimate computer codes are capable of providing more realistic information on the status of the plant, allowing the prediction of real safety margins. The best estimate plus uncertainty approach has proven to be reliable and viable of supplying realistic results if all conditions are carefully followed. This paper, therefore, presents this concept and its possible application to research reactor safety analysis. The aim of the paper is to investigate the unprotected loss-of-flow transients "core blockage" of a miniature neutron source research reactor by applying best estimate plus uncertainty methodology. The results of our calculations show that the temperatures in the core are within the safety limits and do not pose any significant threat to the reactor, as far as the melting of the cladding is concerned. The work also discusses the methodology of the best estimate plus uncertainty approach when applied to the safety analysis of research reactors for licensing purposes.

  16. Application of coupled code technique to a safety analysis of a standard MTR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hamidouche, Tewfik [Division de l' Environnement, de la Surete et des Dechets Radioactifs, Centre de Recherche Nucleaire d' Alger (CRNA), Alger (Algeria); Laboratoire de Mecanique des Fluides Theorique et Appliquee, Faculte de Physique, Universite Des Sciences et de la Technologie Houari Boumediene, (USTHB), Bab-Ezzouar, Alger (Algeria)], E-mail: t.hamidouche@crna.dz; Bousbia-Salah, Anis [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione-Facolta di Ingegneria, Universita di Pisa, Pisa (Italy)], E-mail: b.salah@ing.unipi.it; Si-Ahmed, El Khider [Laboratoire de Mecanique des Fluides Theorique et Appliquee, Faculte de Physique, Universite Des Sciences et de la Technologie Houari Boumediene, (USTHB), Bab-Ezzouar, Alger (Algeria)], E-mail: esi-ahmed@usthb.dz; Mokeddem, Mohamed Yazid [Division de la Physique et des Applications Nucleaires, Centre de Recherche Nucleaire de Draria (CRND) (Algeria); D' Auria, Franscesco [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione-Facolta di Ingegneria, Universita di Pisa, Pisa (Italy)

    2009-10-15

    Accident analyses in nuclear research reactors have been performed, up to now, using simple computational tools based on conservative physical models. These codes, developed to focus on specific phenomena in the reactor, were widely used for licensing purposes. Nowadays, the advances in computer technology make it possible to switch to a new generation of computational tools that provides more realistic description of the phenomena occurring in a nuclear research reactor. Recent International Atomic Energy Agency (IAEA) activities have emphasized the maturity in using Best Estimate (BE) Codes in the analysis of accidents in research reactors. Indeed, some assessments have already been performed using BE thermal-hydraulic system codes such as RELAP5/Mod3. The challenge today is oriented to the application of coupled code techniques for research reactors safety analyses. Within the framework of the current study, a Three-Dimensional Neutron Kinetics Thermal-Hydraulic Model (3D-NKTH) based on coupled PARCS and RELAP5/Mod3.3 codes has been developed for the IAEA High Enriched Uranium (HEU) benchmark core. The results of the steady state calculations are sketched by comparison to tabulated results issued from the IAEA TECDOC 643. These data were obtained using conventional diffusion codes as well as Monte Carlo codes. On the other hand, the transient analysis was assessed with conventional coupled point kinetics-thermal-hydraulic channel codes such as RELAP5 stand alone, RETRAC-PC, and PARET codes. Through this study, the applicability of the coupled code technique is emphasized with an outline of some remaining challenges.

  17. Light-water-reactor safety research program: quarterly progress report, July--September 1977

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-01-01

    Progress is summarized on the Argonne National Laboratory work performed during July, August, and September 1977 on water-reactor-safety problems. The following research and development areas are covered: (1) loss-of-coolant accident research: heat transfer and fluid dynamics; (2) transient fuel response and fission-product release program; (3) mechanical properties of Zircaloy containing oxygen; and (4) steam-explosion studies.

  18. Nodalization effects on RELAP5 results related to MTR research reactor transient scenarios

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2005-01-01

    Full Text Available The present work deals with the anal y sis of RELAP5 results obtained from the evaluation study of the total loss of flow transient with the deficiency of the heat removal system in a research reactor using two different nodalizations. It focuses on the effect of nodalization on the thermal-hydraulic evaluation of the re search reactor. The analysis of RELAP5 results has shown that nodalization has a big effect on the predicted scenario of the postulated transient. There fore, great care should be taken during the nodalization of the reactor, especially when the avail able experimental or measured data are insufficient for making a complete qualification of the nodalization. Our analysis also shows that the research reactor pool simulation has a great effect on the evaluation of natural circulation flow and on other thermal-hydraulic parameters during the loss of flow transient. For example, the on set time of core boiling changes from less than 2000 s to 15000 s, starting from the beginning of the transient. This occurs if the pool is simulated by two vertical volumes in stead of one vertical volume.

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

    Science.gov (United States)

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

    2013-01-01

    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. 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 the room's level where shall be located the rack containing the set sample/detector/shielding. The evaluation of the thermal and epithermal neutron flux in the sample position was done considering the experimental data obtained from a vertical neutron guide, already existent in the reactor, and the simulated model for the facility. The experimental determination of the neutron flux was obtained through the standard procedure of using Au monitors in different positions of the vertical tube. In order to validate both, this experiment and calculations of the simulated model, the flux was also determined in different positions in the core used for sample irradiation. The model of the system was developed using the Monte Carlo code MCNP5. The preliminary results suggest the possibility of obtaining a beam with minimum thermal flux of magnitude 10(6) cm(-2) s(-1), which confirm the technical feasibility of the installation of PGAA at the TRIGA IPR-R1 reactor. This beam would open new possibilities for enhancing the applications using the reactor.

  20. U-Mo Monolithic Fuel for Nuclear Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Prabhakaran, Ramprashad

    2017-11-02

    The metallic fuel selected to replace the current HEU fuels in the research and test reactors is the LEU-10 weight % Mo alloy in the form of a thin sheet or foil encapsulated in AA6061 aluminum alloy with a zirconium interlayer. In order to effectively lead this pursuit, new developments in processing and fabrication of the fuel elements have been initiated, along with a better understanding of material behavior before and after irradiation as a result of these new developments. This editorial note gives an introduction about research and test reactors, need for HEU to LEU conversion, fuel requirements, high uranium density monolithic fuel development and an overview of the four articles published in the December 2017 issue of JOM under a special topic titled “U-Mo Monolithic Fuel for Nuclear Research and Test Reactors”.

  1. Recent upgrades and new scientific infrastructure of MARIA research reactor, Otwock-Swierk, Poland

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-07-01

    The MARIA reactor is open-pool type, water and beryllium moderated. It has two independent primary cooling systems: fuel and pool cooling system. Each fuel assembly is cooled down separately in pressurized channels with individual performances characterization. The fuel assemblies consist of five layers of bent plates or six concentric tubes. Currently it is one of the most powerful research reactors in Europe with operation availability at least up to 2030. Its nominal thermal power is 30 MW. It is characterized by high neutron flux density: up to 3x10{sup 14} n cm{sup -2} s{sup -1} in case of thermal neutrons, and up to 2x10{sup 13} n cm{sup -2} s{sup -1} in case of fast neutrons. The reactor is operated for ca. 4000 h per year. The reactor facility is equipped with fully equipped three hot cells with shielding up to 10{sup 15} Bq. Adjacent to the reactor facility, the radio-pharmaceutics plant (POLATOM) and Material Research Laboratory are located. They are equipped with a number of hot cells with instrumentation. The transport system of radioactive materials from reactor facility to Material Research Laboratory is available. During 2014 the MARIA reactor has been operated with three different types of fuel the same time: previous 36% enriched fuel, and two types of new LEU fuels. In the meantime, molybdenum irradiation programme has been developed. Maria is a multifunctional research tool, with a notable application in production of radioisotopes, radio-pharmaceutics manufacturing (ca. 600 TBq/y), {sup 99}Mo for medical scintigraphy (ca. 6000 TBq/y), neutron transmutation doping of silicon single crystals, wide scientific research based on neutron beams utilization. From the beginning MARIA reactor was intended for loop and fuel testing research activities. Currently it is used mostly as material testing and irradiation facility and for that reason it has wide experimental capabilities. There are eight horizontal irradiation channels from among whom six of them

  2. The International Science and Technology Center (ISTC) and ISTC projects related to research reactors: information review

    Energy Technology Data Exchange (ETDEWEB)

    Tocheniy, L. V.; Rudneva, V. Ya. [ISTC, Moscow (Russian Federation)

    1998-07-01

    The ISTC is an intergovernmental organization established by agreement between the Russian Federation, the European Union, Japan, and the United States. Since 1994, Finland, Sweden, Norway, Georgia, Belarus, Kazakhstan and the Kyrgyz Republic have acceded to the Agreement and Statute. At present, the Republic of Korea is finishing the process of accession to the ISTC. All work of the ISTC is aimed at the goals defined in the ISTC Agreement: To give CIS weapons scientists, particularly those who possess knowledge and skills related to weapons of mass destruction and their delivery systems, the opportunities to redirect their talents to peaceful activities; To contribute to solving national and international technical problems; To support the transition to market-based economics; To support basic and applied research; To help integrate CIS weapons scientists into the international scientific community. The projects may be funded both through governmental funds of the Funding partners of the ISTC. According to the ISTC Statute, approved by the appropriate national organizations, funds used within ISTC projects are exempt from CIS taxes. As of March 1998, more than 1500 proposals had been submitted to the Center, of which 551 were approved for funding, for a total value of approximately US$166 million. The number of scientists and engineers participating in the projects is more than 18000. There are about 20 funded and as yet nonfunded projects related to various problems of research reactors. Many of them address safety issues. Information review of the results and plans of both ongoing projects and as yet nonfunded proposals related to research reactors will be presented with the aim assisting international researchers to establish partnerships or collaboration with ISTC projects. The following groups of ISTC projects will be represented: 1. complex computer simulator s for research reactors; 2. reactor facility decommissioning; 3. neutron sources for medicine; 4

  3. Characterization and quantification of an in-core neutron irradiation facility at a TRIGA II research reactor

    Science.gov (United States)

    Aghara, Sukesh; Charlton, William

    2006-07-01

    Experiments have been performed to characterize the neutron environment at an in-core TRIGA type nuclear research reactor. Steady-state thermal and epithermal neutron environment testing is important for many applications including, materials, electronics and biological cells. A well characterized neutron environment at a research reactor, including energy spectrum and spatial distribution, can be useful to many research communities and for educational research. This paper describes the characterization process and an application of exposing electronics to high neutron fluence.

  4. Design and implementation progress of multi-purpose simulator for nuclear research reactor using LabVIEW

    Energy Technology Data Exchange (ETDEWEB)

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim [Atomic Energy Authority, Cairo (Egypt). Radiation Engineering Dept.; Ashoub, Nagieb [Atomic Energy Authority, Cairo (Egypt). Nuclear Research Center

    2015-11-15

    This paper illustrates the neutronic and thermal hydraulic models that were implemented in the nuclear research reactor simulator based on LabVIEW. It also describes the system and transient analysis of the simulator that takes into consideration the temperature effects and poisoning. This simulator is designed to be a multi-purpose in which the operator could understand the effects of the input parameters on the reactor. A designer can study different solutions for virtual reactor accident scenarios. The main features of the simulator are the flexibility to design and maintain the interface and the ability to redesign and remodel the reactor core engine. The developed reactor simulator permits to acquire hands-on the experience of the physics and technology of nuclear reactors including reactivity control, thermodynamics, technology design and safety system design. This simulator can be easily customizable and upgradable and new opportunities for collaboration between academic groups could be conducted.

  5. Gas Cooled Fast Reactor Research and Development in the European Union

    Directory of Open Access Journals (Sweden)

    Richard Stainsby

    2009-01-01

    Full Text Available Gas-cooled fast reactor (GFR research is directed towards fulfilling the ambitious goals of Generation IV (Gen IV, that is, to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. The research is directed towards developing the GFR as an economic electricity generator, with good safety and sustainability characteristics. Fast reactors maximise the usefulness of uranium resources by breeding plutonium and can contribute to minimising both the quantity and radiotoxicity nuclear waste by actinide transmutation in a closed fuel cycle. Transmutation is particularly effective in the GFR core owing to its inherently hard neutron spectrum. Further, GFR is suitable for hydrogen production and process heat applications through its high core outlet temperature. As such GFR can inherit the non-electricity applications that will be developed for thermal high temperature reactors in a sustainable manner. The Euratom organisation provides a route by which researchers in all European states, and other non-European affiliates, can contribute to the Gen IV GFR system. This paper summarises the achievements of Euratom's research into the GFR system, starting with the 5th Framework programme (FP5 GCFR project in 2000, through FP6 (2005 to 2009 and looking ahead to the proposed activities within the 7th Framework Programme (FP7.

  6. Source term derivation and radiological safety analysis for the TRICO II research reactor in Kinshasa

    Energy Technology Data Exchange (ETDEWEB)

    Muswema, J.L., E-mail: jeremie.muswem@unikin.ac.cd [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Ekoko, G.B. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Lukanda, V.M. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Democratic Republic of the Congo' s General Atomic Energy Commission, P.O. Box AE1 (Congo, The Democratic Republic of the); Lobo, J.K.-K. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Darko, E.O. [Radiation Protection Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana); Boafo, E.K. [University of Ontario Institute of Technology, 2000 Simcoe St. North, Oshawa, ONL1 H7K4 (Canada)

    2015-01-15

    Highlights: • Atmospheric dispersion modeling for two credible accidents of the TRIGA Mark II research reactor in Kinshasa (TRICO II) was performed. • Radiological safety analysis after the postulated initiating events (PIE) was also carried out. • The Karlsruhe KORIGEN and the HotSpot Health Physics codes were used to achieve the objectives of this study. • All the values of effective dose obtained following the accident scenarios were below the regulatory limits for reactor staff members and the public, respectively. - Abstract: The source term from the 1 MW TRIGA Mark II research reactor core of the Democratic Republic of the Congo was derived in this study. An atmospheric dispersion modeling followed by radiation dose calculation were performed based on two possible postulated accident scenarios. This derivation was made from an inventory of peak radioisotope activities released in the core by using the Karlsruhe version of isotope generation code KORIGEN. The atmospheric dispersion modeling was performed with HotSpot code, and its application yielded to radiation dose profile around the site using meteorological parameters specific to the area under study. The two accident scenarios were picked from possible accident analyses for TRIGA and TRIGA-fueled reactors, involving the case of destruction of the fuel element with highest activity release and a plane crash on the reactor building as the worst case scenario. Deterministic effects of these scenarios are used to update the Safety Analysis Report (SAR) of the reactor, and for its current version, these scenarios are not yet incorporated. Site-specific meteorological conditions were collected from two meteorological stations: one installed within the Atomic Energy Commission and another at the National Meteorological Agency (METTELSAT), which is not far from the site. Results show that in both accident scenarios, radiation doses remain within the limits, far below the recommended maximum effective

  7. Material characteristics and construction methods for a typical research reactor concrete containment in Iran

    Energy Technology Data Exchange (ETDEWEB)

    Ebrahimia, Mahsa; Suha, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of); Eghbalic, Rahman; Jahan, Farzaneh Asadi malek [School of Architecture and Urbanism, Qazvin (Iran, Islamic Republic of)

    2012-10-15

    Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran.

  8. The neutronic analysis of opportunity of ITER blanket element tests in RF research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lopatkin, A.; Tocheniy, L. [ENTEK-RDIPE, Moscow (Russian Federation)

    1994-12-31

    In the framework of development of plan of in-pile radiative tests of ITER blanket elements the calculations are carried out of the models of tritium-producing elements in loop channels, placed in the number of Russian various type test reactors. There are presented: (1) The variants of models of blanket, on the base of which the set of experiment goal parameters and its ranges are formed; (2) Outline of loop channel; (3) The experimental opportunities of research reactors with thermal (SM-3, MIR, IVV-2M, RBT) and fast (BOR-60, EBR) spectra of neutrons; (4) The calculation procedures - settlement models, codes. The results are given: (1) power generation rates in components of channel; (2) the tritium breeding rate; (3) the helium production rate in beryllium; (4) the neutron group fluxes; (5) absorption rates in zones of loop channel. The possible reactivity effects due to experimental channel accommodation in reactor core and to radiated sample replace inside of the channel are shown. The last section includes the recommendations for the choice of reactor acceptable from the neutronics point of view, and for the next study directions and stages.

  9. Evaluation of Gamma Fluence Rate Predictions for 41-argon Releases to the Atmosphere at a Nuclear Research Reactor Site

    DEFF Research Database (Denmark)

    Rojas-Palma, Carlos; Aage, Helle Karina; Astrup, Poul

    2004-01-01

    An experimental study of radionuclide dispersion in the atmosphere has been conducted at the BR1 research reactor in Mol, Belgium. Artificially generated aerosols ('white smoke') were mixed with the routine releases of Ar-41 in the reactor's 60-m tall venting stack. The detailed plume geometry...

  10. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

    Science.gov (United States)

    Lashkari, A.; Khalafi, H.; Kazeminejad, H.

    2013-01-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

  11. Nuclear Reactors. Revised.

    Science.gov (United States)

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

  12. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

  13. IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

    Energy Technology Data Exchange (ETDEWEB)

    Yamada, K. [Vienna International Centre, P.O. Box 100, 1400 Vienna (Austria); Aksan, S. N. [International Atomic Energy Agency, 1400 Vienna (Austria)

    2012-07-01

    The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present, 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)

  14. Progress report on neutron activation analysis at Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tuan, Nguyen Ngoc [Nuclear Research Institute, Dalat (Viet Nam)

    2003-03-01

    Neutron Activation Analysis (NAA) is one of most powerful techniques for the simultaneous multi-elements analysis. This technique has been studied and applied to analyze major, minor and trace elements in Geological, Biological and Environmental samples at Dalat Nuclear Research Reactor. At the sixth Workshop, February 8-11, 1999, Yojakarta, Indonesia we had a report on Current Status of Neutron Activation Analysis using Dalat Nuclear Research Reactor. Another report on Neutron Activation Analysis at the Dalat Nuclear Research Reactor also was presented at the seventh Workshop in Taejon, Korea from November 20-24, 2000. So in this report, we would like to present the results obtained of the application of NAA at NRI for one year as follows: (1) Determination of the concentrations of noble, rare earth, uranium, thorium and other elements in Geological samples according to requirement of clients particularly the geologists, who want to find out the mineral resources. (2) The analysis of concentration of radionuclides and nutrient elements in foodstuffs to attend the program on Asian Reference Man. (3) The evaluation of the contents of trace elements in crude oil and basement rock samples to determine original source of the oil. (4) Determination of the elemental composition of airborne particle in the Ho Chi Minh City for studying air pollution. The analytical data of standard reference material, toxic elements and natural radionuclides in seawater are also presented. (author)

  15. Available reprocessing and recycling services for research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tozser, Sandor Miklos; Adelfang, Pablo; Bradley, Ed [International Atomic Energy Agency, Vienna (Austria); Budu, Madalina [SOSNY Research and Development Company, Moscow (Russian Federation); Chiguer, Mustapha [AREVA, Paris (France)

    2015-05-15

    International activities in the back-end of the research reactor (RR) fuel cycle have so far been dominated by the programmes of acceptance of highly-enriched uranium (HEU) spent nuclear fuel (SNF) by the country where it was originally enriched. These programmes will soon have achieved their goals and the SNF take-back programmes will cease. However, the needs of the nuclear community dictate that the majority of the research reactors continue to operate using low enriched uranium (LEU) fuel in order to meet the varied mission objectives. As a result, inventories of LEU SNF will continue to be created and the back-end solution of RR SNF remains a critical issue. In view of this fact, the IAEA, based on the experience gained during the decade of international cooperation in supporting the objectives of the HEU take-back programmes, will draw up a report presenting available reprocessing and recycling services for research reactor spent nuclear fuel. This paper gives an overview of the guiding document which will address all aspects of Reprocessing and Recycling Services for RR SNF, including an overview of solutions, decision making support, service suppliers, conditions (prerequisites, options, etc.), services offered by the managerial and logistics support providers with a focus on available transport packages and applicable transport modes.

  16. Event and fault tree model for reliability analysis of the greek research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Albuquerque, Tob R.; Guimaraes, Antonio C.F.; Moreira, Maria de Lourdes, E-mail: atalbuquerque@ien.gov.br, E-mail: btony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    Fault trees and event trees are widely used in industry to model and to evaluate the reliability of safety systems. Detailed analyzes in nuclear installations require the combination of these two techniques. This work uses the methods of fault tree (FT) and event tree (ET) to perform the Probabilistic Safety Assessment (PSA) in research reactors. The PSA according to IAEA (International Atomic Energy Agency) is divided into Level 1, Level 2 and level 3. At Level 1, conceptually safety systems act to prevent the accident, at Level 2, the accident occurred and seeks to minimize the consequences, known as stage management of the accident, and at Level 3 are determined consequences. This paper focuses on Level 1 studies, and searches through the acquisition of knowledge consolidation of methodologies for future reliability studies. The Greek Research Reactor, GRR - 1, was used as a case example. The LOCA (Loss of Coolant Accident) was chosen as the initiating event and from there were developed the possible accident sequences, using event tree, which could lead damage to the core. Furthermore, for each of the affected systems, the possible accidents sequences were made fault tree and evaluated the probability of each event top of the FT. The studies were conducted using a commercial computational tool SAPHIRE. The results thus obtained, performance or failure to act of the systems analyzed were considered satisfactory. This work is directed to the Greek Research Reactor due to data availability. (author)

  17. NUMERICAL SIMULATION FOR MECHANICAL BEHAVIOR OF U10MO MONOLITHIC MINIPLATES FOR RESEARCH AND TEST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Hakan Ozaltun & Herman Shen

    2011-11-01

    This article presents assessment of the mechanical behavior of U-10wt% Mo (U10Mo) alloy based monolithic fuel plates subject to irradiation. Monolithic, plate-type fuel is a new fuel form being developed for research and test reactors to achieve higher uranium densities within the reactor core to allow the use of low-enriched uranium fuel in high-performance reactors. Identification of the stress/strain characteristics is important for understanding the in-reactor performance of these plate-type fuels. For this work, three distinct cases were considered: (1) fabrication induced residual stresses (2) thermal cycling of fabricated plates; and finally (3) transient mechanical behavior under actual operating conditions. Because the temperatures approach the melting temperature of the cladding during the fabrication and thermal cycling, high temperature material properties were incorporated to improve the accuracy. Once residual stress fields due to fabrication process were identified, solution was used as initial state for the subsequent simulations. For thermal cycling simulation, elasto-plastic material model with thermal creep was constructed and residual stresses caused by the fabrication process were included. For in-service simulation, coupled fluid-thermal-structural interaction was considered. First, temperature field on the plates was calculated and this field was used to compute the thermal stresses. For time dependent mechanical behavior, thermal creep of cladding, volumetric swelling and fission induced creep of the fuel foil were considered. The analysis showed that the stresses evolve very rapidly in the reactor. While swelling of the foil increases the stress of the foil, irradiation induced creep causes stress relaxation.

  18. Reactor safety research programs. Quarterly progress report, January 1--March 31, 1977

    Energy Technology Data Exchange (ETDEWEB)

    Romano, A.J. (comp.)

    1977-05-01

    The projects reported each quarter are the following: Gas Reactor Safety Evaluation, THOR Code Development, SSC Code Development, LMFBR and LWR Safety Experiments, Fast Reactor Safety Code Validation, Technical Coordination of Structural Integrity, and Fast Reactor Safety Reliability Assessment.

  19. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    Energy Technology Data Exchange (ETDEWEB)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials

  20. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have

  1. In-Vessel Melt Retention of Pressurized Water Reactors: Historical Review and Future Research Needs

    Directory of Open Access Journals (Sweden)

    Weimin Ma

    2016-03-01

    Full Text Available A historical review of in-vessel melt retention (IVR is given, which is a severe accident mitigation measure extensively applied in Generation III pressurized water reactors (PWRs. The idea of IVR actually originated from the back-fitting of the Generation II reactor Loviisa VVER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse AP1000, the Korean APR1400 as well as Chinese advanced PWR designs HPR1000 and CAP1400. The most influential phenomena on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the vessel by the molten core, and the external cooling of the reactor pressure vessel (RPV. For in-vessel melt evolution, past focus has only been placed on the melt pool convection in the lower plenum of the RPV; however, through our review and analysis, we believe that other in-vessel phenomena, including core degradation and relocation, debris formation, and coolability and melt pool formation, may all contribute to the final state of the melt pool and its thermal loads on the lower head. By looking into previous research on relevant topics, we aim to identify the missing pieces in the picture. Based on the state of the art, we conclude by proposing future research needs.

  2. IGORR-1: Proceedings of the first meeting of the international group on research reactors

    Energy Technology Data Exchange (ETDEWEB)

    West, C.D. (comp.)

    1990-05-01

    Many organizations, in several countries, are planning or implementing new or upgraded research reactor projects, but there has been no organized forum devoted entirely to discussion and exchange of information in this field. Over the past year or so, informal discussions resulted in widespread agreement that such a forum would serve a useful purpose. Accordingly, a proposal to form a group was submitted to the leading organizations known to be involved in projects to build or upgrade reactor facilities. Essentially all agreed to join in the formation of the International Group on Research Reactors (IGORR) and nominated a senior staff member to serve on its international organizing committee. The first IGORR meeting took place on February 28--March 2, 1990. It was very successful and well attended; some 52 scientists and engineers from 25 organizations in 10 countries participated in 2-1/2 days of open and informative presentations and discussions. Two workshop sessions offered opportunities for more detailed interaction among participants and resulted in identification of common R D needs, sources of data, and planned new facilities. Individual papers have been cataloged separately.

  3. OVERVIEW OF CRITERIA FOR INTERIM WET & DRY STORAGE OF RESEARCH REACTOR SPENT NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.; Vinson, D.; Iyer, N.; Fisher, D.

    2010-11-03

    Following discharge from research reactors, spent nuclear fuel may be stored 'wet' in water pools or basins, or it may be stored 'dry' in various configurations including non-sealed or sealed containers until retrieved for ultimate disposition. Interim safe storage practices are based on avoiding degradation to the fuel that would impact functions related to safety. Recommended practices including environmental controls with technical bases, are outlined for wet storage and dry storage of aluminum-clad, aluminum-based research reactor fuel. For wet storage, water quality must be maintained to minimize corrosion degradation of aluminum fuel. For dry storage, vented canister storage of aluminum fuel readily provides a safe storage configuration. For sealed dry storage, drying must be performed so as to minimize water that would cause additional corrosion and hydrogen generation. Consideration must also be given to the potential for radiolytically-generated hydrogen from the bound water in the attendant oxyhydroxides on aluminum fuel from reactor operation for dry storage systems.

  4. Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

    Science.gov (United States)

    Clement, J. D.; Rust, J. H.; Schneider, A.; Hohl, F.

    1976-01-01

    The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

  5. Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-01

    The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

  6. Analysis of neutron flux distribution for the validation of computational methods for the optimization of research reactor utilization.

    Science.gov (United States)

    Snoj, L; Trkov, A; Jaćimović, R; Rogan, P; Zerovnik, G; Ravnik, M

    2011-01-01

    In order to verify and validate the computational methods for neutron flux calculation in TRIGA research reactor calculations, a series of experiments has been performed. The neutron activation method was used to verify the calculated neutron flux distribution in the TRIGA reactor. Aluminium (99.9 wt%)-Gold (0.1 wt%) foils (disks of 5mm diameter and 0.2mm thick) were irradiated in 33 locations; 6 in the core and 27 in the carrousel facility in the reflector. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and experimental normalized reaction rates in the core are in very good agreement for both isotopes indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux and reaction rate distribution in the reactor core. In the reflector however, the accuracy of the epithermal and thermal neutron flux distribution and attenuation is lower, mainly due to lack of information about the material properties of the graphite reflector surrounding the core, but the differences between measurements and calculations are within 10%. Since our computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of research reactor utilization. Copyright © 2010 Elsevier Ltd. All rights reserved.

  7. Production and modeling of radioactive gold nanoparticles in Tehran research reactor.

    Science.gov (United States)

    Hosseini, Seyedeh Fatemeh; Sadeghi, Mahdi; Aboudzadeh, Mohammad Reza; Mohseni, Morteza

    2016-12-01

    Gold has two medically useful radioactive isotopes, 198Au and 199Au, for locally irradiating and killing tumor cells. 198Au radionuclide has been produced through the irradiation of the pure gold via 197Au(n,γ)198Au reaction in the Tehran Research Reactor at a thermal neutron flux of 4.5×1013ncm-2s-1 for the different irradiation times. In this paper, the activity of 198Au radionuclide has been determined using MCNPX-2.6 and TALYS-1.6 codes and also the theoretical approach. The calculated results were compared with the corresponding experimental values. The calculated results were in good agreement with the experimental data, thus the used codes can be used as a powerful tool to predict and optimize production conditions in reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

  8. Releasable activity and maximum permissible leakage rate within a transport cask of Tehran Research Reactor fuel samples

    National Research Council Canada - National Science Library

    Rezaeian Mahdi; Kamali Jamshid; Roshanzamir Manoochehr; Moosakhani Alireza; Noori Elghar

    2015-01-01

    ... operators. Based on IAEA regulations, releasable activity and maximum permissible volumetric leakage rate within the cask containing fuel samples of Tehran Research Reactor enclosed in an irradiated capsule are calculated...

  9. Decommissioning of the ASTRA research reactor: Planning, executing and summarizing the project

    Directory of Open Access Journals (Sweden)

    Meyer Franz

    2010-01-01

    Full Text Available The decommissioning of the ASTRA research reactor at the Austrian Research Centres Seibersdorf was described within three technical papers already released in Nuclear Technology & Radiation Protection throughout the years 2003, 2006, and 2008. Following a suggestion from IAEA the project was investigated well after the files were closed regarding rather administrative than technical matters starting with the project mission, explaining the project structure and identifying the key factors and the key performance indicators. The continuous documentary and reporting system as implemented to fulfil the informational needs of stake-holders, management, and project staff alike is described. Finally the project is summarized in relationship to the performance indicators.

  10. The Operational Parameter Electronic Database of the IPR-R1 TRIGA Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Zacarias; Gomes do Prado Souza, Rose Mary [Nuclear Technology Development Center (CDTN), Belo Horizonte - MG (Brazil)

    2008-10-29

    it in a historical database. In every operation of the research reactor about forty variables are registered by the data acquisition system. The system provides data for the last five years (since 2004)

  11. A world class nuclear research reactor complex for South Africa's nuclear future

    Energy Technology Data Exchange (ETDEWEB)

    Keshaw, Jeetesh [South African Young Nuclear Professional Society, PO Box 9396, Centurion, 0157 (South Africa)

    2008-07-01

    South Africa recently made public its rather ambitious goals pertaining to nuclear energy developments in a Draft Policy and Strategy issued for public comment. Not much attention was given to an important tool for nuclear energy research and development, namely a well equipped and maintained research reactor, which on its own does not do justice to its potential, unless it is fitted with all the ancillaries and human resources as most first world countries have. In South Africa's case it is suggested to establish at least one Nuclear Energy Research and Development Centre at such a research reactor, where almost all nuclear energy related research can be carried out on par with some of the best in the world. The purpose of this work is to propose how this could be done, and motivate why it is important that it be done with great urgency, and with full involvement of young professionals, if South Africa wishes to face up to the challenges mentioned in the Draft Strategy and Policy. (authors)

  12. RESEARCHES ON THE DIGESTERS AND REACTORS WHICH CAN BE USED IN A FARM SCALE BIOGAS PLANT

    Directory of Open Access Journals (Sweden)

    Mariana DUMITRU

    2014-10-01

    Full Text Available In the general context of searching integrated system of renewable energy production, this paper present some researches on the reactors and the digesters, as a main part of a biogas plant at a farm scale. After we present the most used types of digesters, we also concentrated over the processes which take place into a digester, one of them being the removal of H2S from biogas (desulphurisation, which can be made by various methods, either biological or chemical, taking place inside or outside the digester. In the case of biological desulphurization outside the digester, we concentrate on the types of reactors which can be used in this case. Beside the well known types of reactors, we present the possibility of using an original self pressure membrane bioreactor. In this type of bioreactor, the metabolic activity of gas producing microorganisms, especially yeast, could obtain high pressure from gas produced in closed medium on the one hand, and separation of other products of metabolism through membrane on the other hand, using gas pressure as driving force. It is known that several strains of yeast resist on very high hydrostatic pressure heaving good activity. This fact give the possibility to use their energy for other purposes, such as producing mechanical work. Combination of both, gas pressure and alchool burning, increase the process efficiency.

  13. 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.

  14. 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. Copyright © 2015 Elsevier Ltd. All rights reserved.

  15. Power up-grading study for the first Egyptian research reactor

    Energy Technology Data Exchange (ETDEWEB)

    El-Sawy Temraz, H.; Ashoub, N. E-mail: nageeb@pcn.aea.sci.eg; Fathallah, A

    2001-09-01

    In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4x4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5x5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2-10 MW) and different core coolant flow rates (450, 900, 1350 m{sup 3} h{sup -1}) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5-group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4x4-fuel basket and with the proposed 5x5-fuel basket up to 5 MW with the present coolant flow rate (900 m{sup 3} h{sup -1}). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m{sup 3} h{sup -1} without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5x5) increases the excess reactivity of the reactor core than the present 4x4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.

  16. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  17. Complementary Safety Margin Assessment f the Nuclear Installations of the research reactor in Petten, Netherlands

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-02-15

    On March 11, 2011, a large part of the Japanese eastern coastal area was devastated by an earthquake, followed by an immense tsunami. As a result, thousands of people were killed, injured or made homeless. In the days that followed, the situation was further complicated because of the failing nuclear reactors on the Fukushima coast. The local environment suffered from radioactive releases, requiring evacuation zones, and generating international concerns about nuclear safety. In the wake of this disaster the European Union decided to assess safety on all operating nuclear reactors in its member states. This safety evaluation initiated by the European Union focusses on extreme natural hazards, beyond the standard safety evaluations which regularly have to be performed to demonstrate the safety of a nuclear power plant. Consequences of these extreme hazards for the research reactor in Petten, Netherlands, have been evaluated based on available safety analyses, supplemented by engineering judgement. In this way, the robustness of the existing plant has been assessed and possible measures to further increase the safety margins have been identified. This document presents the results of the Complementary Safety margin Assessment (CSA) performed for the 'Onderzoekslocatie Petten'. The distinct difference between this report and former risk analysis reports in general and the existing Safety Report of the Petten reactor is that the maximum resistance of the plant against redefined and more challenging events has been investigated, whereas traditionally the plant design is investigated against certain events that are determined on a historical basis. This different approach requires different analyses and studies, which in turn presents new insights into the robustness of the plant. The main purpose of this report is to answer the questions posed by the Ministry of Economic Affairs, Agriculture and Innovation. It was decided to write at the same time a report in

  18. Concept of a nuclear powered submersible research vessel and a compact reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Takahashi, Teruo [Energis, Co., Kobe, Hyogo (Japan); Nishimura, Hajime [Japan Marine Science and Technology Center, Yokosuka, Kanagawa (Japan); Tokunaga, Sango [Japan Deep Sea Technology Association, Tokyo (Japan)

    2001-07-01

    A conceptual design study of a submersible research vessel navigating in 600 m depth and a compact nuclear reactor were carried out for the expansion of the nuclear power utilization. The mission of the vessel is the research of mechanism of the climate change to predict the global environment. Through conditions of the Arctic Ocean and the sea at high latitude have significant impacts on the global environmental change, it is difficult to investigate those areas by ordinary ships because of thick ice or storm. Therefore the research vessel is mainly utilized in the Arctic Ocean and the sea at high latitude. By taking account of the research mission, the basic specifications of the vessel are decided; the total weight is 500 t, the submersible depth is 600 m, the maximum speed is 12 knots (22.2 km/h), and the number of crews is 16. Nuclear power has an advantage in supplying large power of electricity in the sea for long period. Based on the requirements, it has been decided that two sets of submersible compact reactor, SCR, which is light-weighted and of enhanced safety characteristics of supply the total electricity of 500 kW. (author)

  19. NCTPlan application for neutron capture therapy dosimetric planning at MEPhI nuclear research reactor.

    Science.gov (United States)

    Elyutina, A S; Kiger, W S; Portnov, A A

    2011-12-01

    The results of modeling of two therapeutic beams HEC-1 and HEC-4 at the NRNU "MEPhI" research nuclear reactor exploitable for preclinical treatments are reported. The exact models of the beams are constructed as an input to the NCTPlan code used for planning Neutron Capture Therapy (NCT) procedure. The computations are purposed to improve the accuracy of prediction of a dose absorbed in tissue with the account of all components of radiation. Copyright © 2011 Elsevier Ltd. All rights reserved.

  20. Determination of fast neutron flux distribution in irradiation sites of the Malaysian Nuclear Agency research reactor.

    Science.gov (United States)

    Yavar, A R; Sarmani, S B; Wood, A K; Fadzil, S M; Radir, M H; Khoo, K S

    2011-05-01

    Determination of thermal to fast neutron flux ratio (f(fast)) and fast neutron flux (ϕ(fast)) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f(fast) and subsequently ϕ(fast) were determined using the absolute method. The f(fast) ranged from 48 to 155, and the ϕ(fast) was found in the range 1.03×10(10)-4.89×10(10) n cm(-2) s(-1). These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor. Copyright © 2011 Elsevier Ltd. All rights reserved.

  1. Decontamination and decommissioning project of the TRIGA mark - 2 and 3 research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jung, K. J.; Baik, S. T.; Chung, U. S.; Jung, K. H.; Park, S. K.; Kim, J. K.; Lee, D. G.; Kim, H. R.; Lee, B. J.; Yang, S. H.

    2001-01-15

    The decommissioning license for KRR (Korea Research Reactor) 1 and 2 was issued Nov. 23, 2000. The atmospheric stability on the KRR site was evaluated using the meteorological data measured at the site. From the results of this evaluation, the population dose was evaluated for the public who lives at the periphery of the site. The Radiation Safety Management Guideline was developed and it will be used as a base line making Radiation Safety Management Procedure. The container was specially designed and manufactured for the storing of low level radioactive solid waste arising from the D and D activities. Firstly, the 50 containers were completely manufactured.

  2. Replacement Nuclear Research Reactor: Draft Environmental Impact Statement. Vol. 2. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-07-01

    The appendices contains additional relevant information on: Environment Australia EIS Guidelines, composition of the Study Team, Consultation Activities and Resuits, Relevant Legislation and Regulatory Requirements, Exampies of Multi-Purpose Research Reactors, Impacts of Radioactive Emissions and Wastes Generated at Lucas Heights Science and Technology Centre, Technical Analysis of the Reference Accident, Flora and Fauna Species Lists, Summary of Environmental Commitments and an Outline of the Construction Environmental Management Plan Construction Environmental Management Plan figs., ills., refs. Prepared for Australian Nuclear Science and Technology Organisation (ANSTO)

  3. 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. Copyright © 2015 Elsevier Ltd. All rights reserved.

  4. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

  5. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  6. Safety evaluation report related to the renewal of the facility license for the research reactor at the Dow Chemical Company

    Energy Technology Data Exchange (ETDEWEB)

    1989-04-01

    This safety evaluation report for the application filed by the Dow Chemical Company for renewal of facility Operating License R-108 to continue to operate its research reactor at an increased operating power level has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located on the grounds of the Michigan Division of the Dow Chemical Company in Midland, Michigan. The staff concludes that the Dow Chemical Company can continue to operate its reactor without endangering the health and safety of the public.

  7. Reliability studies in research reactors; Estudo de confiabilidade em reatores de pesquisa

    Energy Technology Data Exchange (ETDEWEB)

    Albuquerque, Tob Rodrigues de

    2013-08-01

    Fault trees and event trees are widely used in industry to model and to evaluate the reliability of safety systems. Detailed analyzes in nuclear installations require the combination of these two techniques. This study uses the methods of FT (Fault Tree) and ET (Event Tree) to accomplish the PSA (Probabilistic Safety Assessment) in research reactors. According to IAEA (lnternational Atomic Energy Agency), the PSA is divided into Level 1, Level 2 and Level 3. At the Level 1, conceptually, the security systems perform to prevent the occurrence of accidents, At the Level 2, once accidents happened, this Level seeks to minimize consequences, known as stage management of accident, and at Level 3 accident impacts are determined. This study focuses on analyzing the Level 1, and searching through the acquisition of knowledge, the consolidation of methodologies for future reliability studies. The Greek Research Reactor, GRR-1, is a case example. The LOCA (Loss of Coolant Accident) was chosen as the initiating event and from it, using ET, possible accidental sequences were developed, which could lead damage to the core. Moreover, for each of affected systems, probabilities of each event top of FT were developed and evaluated in possible accidental sequences. Also, the estimates of importance measures for basic events are presented in this work. The studies of this research were conducted using a commercial computational tool SAPHIRE. Additionally, achieved results thus were considered satisfactory for the performance or the failure of analyzed systems. (author)

  8. Assessment of the reliability of neutronic parameters of Ghana Research Reactor-1 control systems

    Energy Technology Data Exchange (ETDEWEB)

    Amponsah-Abu, E.O., E-mail: edwardabu2002@yahoo.com [National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG. 80, Legon-Accra (Ghana); Gbadago, J.K. [National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG. 80, Legon-Accra (Ghana); Akaho, E.H.K.; Akoto-Bamford, S. [School of Nuclear and Allied Sciences, University of Ghana (Ghana); Gyamfi, K.; Asamoah, M.; Baidoo, I.K. [National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG. 80, Legon-Accra (Ghana)

    2015-01-15

    Highlights: • The reliability of neutronics parameters of GHARR-I was assessed. • The reactor was operated at different power levels of 5–30 kW. • The pre-set flux was compared with the flux in the inner irradiation site. • Decrease in the core reactivity caused difference in flux on the meters and site. • Neutronic parameters become reliable when operation is done at reactivity of 4 mk. - Abstract: The Ghana Research Reactor-1 (GHARR-1) has been in operation for the past 19 years using a Micro-Computer Closed Loop System (MCCLS) and Control Console (CC) as the control systems. The two control systems were each coupled separately with a micro-fission chamber to measure the current pulses of the neutron fluxes in the core at excess reactivity of 4 mk. The MCCLS and CC meter readings at a pre-set flux of 5.0 × 10{sup 11} n/cm{sup 2} s were 6.42 × 10{sup 11} n/cm{sup 2} s and 5.0 × 10{sup 11} n/cm{sup 2} s respectively. Due to ageing and obsolescence, the MCCLS and some components that control the sensitivity and the reading mechanism of the meters were replaced. One of the fission chambers was also removed and the two control systems were coupled to one fission chamber. The reliability of the neutronic parameters of the control systems was assessed after the replacement. The results showed that when the reactor is operated at different power levels of 5–30 kW using one micro-fission chamber, the pre-set neutron fluxes at the control systems is 1.6 times the neutron fluxes obtained using a flux monitor at the inner irradiation site two of the reactor. The average percentage deviations of the obtained fluxes from the pre-set values of 1.67 × 10{sup 11}–1.0 × 10{sup 12} n/cm{sup 2} s were 36.5%. This compares very well with the decrease in core excess reactivity of 36.3% of the nominal value of 4 mk, after operating the reactor at critical neutron flux of 1.0 × 10{sup 9} n/cm{sup 2} s.

  9. Supercritical Water Reactor (SCWR) - Survey of Materials Research and Development Needs to Assess Viability

    Energy Technology Data Exchange (ETDEWEB)

    Philip E. MacDonald

    2003-09-01

    Supercritical water-cooled reactors (SCWRs) are among the most promising advanced nuclear systems because of their high thermal efficiency [i.e., about 45% vs. 33% of current light water reactors (LWRs)] and considerable plant simplification. SCWRs achieve this with superior thermodynamic conditions (i.e., high operating pressure and temperature), and by reducing the containment volume and eliminating the need for recirculation and jet pumps, pressurizer, steam generators, steam separators and dryers. The reference SCWR design in the U.S. is a direct cycle, thermal spectrum, light-water-cooled and moderated reactor with an operating pressure of 25 MPa and inlet/outlet coolant temperature of 280/500 °C. The inlet flow splits, partly to a down-comer and partly to a plenum at the top of the reactor pressure vessel to flow downward through the core in special water rods to the inlet plenum. This strategy is employed to provide good moderation at the top of the core, where the coolant density is only about 15-20% that of liquid water. The SCWR uses a power conversion cycle similar to that used in supercritical fossil-fired plants: high- intermediate- and low-pressure turbines are employed with one moisture-separator re-heater and up to eight feedwater heaters. The reference power is 3575 MWt, the net electric power is 1600 MWe and the thermal efficiency is 44.8%. The fuel is low-enriched uranium oxide fuel and the plant is designed primarily for base load operation. The purpose of this report is to survey existing materials for fossil, fission and fusion applications and identify the materials research and development needed to establish the SCWR viabilitya with regard to possible materials of construction. The two most significant materials related factors in going from the current LWR designs to the SCWR are the increase in outlet coolant temperature from 300 to 500 °C and the possible compatibility issues associated with the supercritical water environment.

  10. The effect of code user and boundary conditions on RELAP calculations of MTR research reactor transient scenarios

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2005-01-01

    Full Text Available The safety evaluation of nuclear power and re search reactors is a very important step before their construction and during their operation. This evaluation based on the best estimate calculations requires qualified codes qualified users, and qualified nodalizations. The effect of code users on the RELAP5 results during the analysis of loss of flow transient in MTR research reactors is presented in this pa per. To clarify this effect, two nodalizations for research reactor different in the simulation of the open water surface boundary conditions of the reactor pool have been used. Very different results are obtained with few choices for code users. The core natural circulation flow with the be ginning of core boiling doesn't stop but in creases. The in creasing in the natural circulation flow shifts out the boiling from the core and the clad temperature decreases be low the local saturation temperature.

  11. Investigation of Classification and Design Requirements for Digital Software for Advanced Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Park, Gee Young; Jung, H. S.; Ryu, J. S.; Park, C

    2005-06-15

    As the digital technology is being developed drastically, it is being applied to various industrial instrumentation and control (I and C) fields. In the nuclear power plants, I and C systems are also being installed by digital systems replacing their corresponding analog systems installed previously. There had been I and C systems constructed by analog technology especially for the reactor protection system in the research reactor HANARO. Parallel to the pace of the current trend for digital technology, it is desirable that all I and C systems including the safety critical and non-safety systems in an advanced research reactor is to be installed based on the computer based system. There are many attractable features in using digital systems against existing analog systems in that the digital system has a superior performance for a function and it is more flexible than the analog system. And any fruit gained from the newly developed digital technology can be easily incorporated into the existing digital system and hence, the performance improvement of a computer based system can be implemented conveniently and promptly. Moreover, the capability of high integrity in electronic circuits reduces the electronic components needed to construct the processing device and makes the electronic board simple, and this fact reveals that the hardware failure itself are unlikely to occur in the electronic device other than some electric problems. Balanced the fact mentioned above are the roles and related issues of the software loaded on the digital integrated hardware. Some defects in the course of software development might induce a severe damage on the computer system and plant systems and therefore it is obvious that comprehensive and deep considerations are to be placed on the development of the software in the design of I and C system for use in an advanced research reactor. The work investigates the domestic and international standards on the classifications of digital

  12. Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

    Energy Technology Data Exchange (ETDEWEB)

    Sekhri, Abdelkrim; Graham, Andy [RRT Radiation and Reactor Theory, South African Nuclear Energy Corporation - NECSA, PO Box 582 Pretoria 0001 (South Africa); D' Arcy, Alan; Oliver, Melissa [SAFARI-1 Research Reactor, South African Nuclear Energy Corporation - NECSA, PO Box 582 Pretoria 0001 (South Africa)

    2008-07-01

    The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

  13. 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...

  14. Special issue on the "Consortium for Advanced Simulation of Light Water Reactors Research and Development Progress"

    Science.gov (United States)

    Turinsky, Paul J.; Martin, William R.

    2017-04-01

    In this special issue of the Journal of Computational Physics, the research and development completed at the time of manuscript submission by the Consortium for Advanced Simulation of Light Water Reactors (CASL) is presented. CASL is the first of several Energy Innovation Hubs that have been created by the Department of Energy. The Hubs are modeled after the strong scientific management characteristics of the Manhattan Project and AT&T Bell Laboratories, and function as integrated research centers that combine basic and applied research with engineering to accelerate scientific discovery that addresses critical energy issues. Lifetime of a Hub is expected to be five or ten years depending upon performance, with CASL being granted a ten year lifetime.

  15. Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P.; Ryu, H. J.; Kulesza, K. P.; Kim, S. J.; Halsey, W. G.; Smith, C. F.; Brown, N. W.; Greenspan, E.; de Caro, M.; Li, N.; Hosemann, P.; Zhang, J.; Yu, H.; Nuclear Engineering Division; LLNL; LANL; Massachusetts Inst. of Tech.; Ecole des Mines de Paris; Oregon State Univ.; Univ.of California at Berkley

    2008-06-23

    This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics

  16. Thermal-Hydraulic Research Review and Cooperation Outcome for Light Water Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Shin, Chang Hwan; Lee, Chan; Chun, Tae Hyun; Oh, Dong Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Chi Young [Pukyong Nat’l Univ., Busan (Korea, Republic of)

    2016-12-15

    The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermalhydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermalhydraulic technology and the commercialization.

  17. Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.

    1999-02-01

    The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses.

  18. The current state of the Russian reduced enrichment research reactors program

    Energy Technology Data Exchange (ETDEWEB)

    Aden, V.G.; Kartashov, E.F.; Lukichev, V.A. [and others

    1997-08-01

    During the last year after the 16-th International Conference on Reducing Fuel Enrichment in Research Reactors held in October, 1993 in Oarai, Japan, the conclusive stage of the Program on reducing fuel enrichment (to 20% in U-235) in research reactors was finally made up in Russia. The Program was started late in 70th and the first stage of the Program was completed by 1986 which allowed to reduce fuel enrichment from 80-90% to 36%. The completion of the Program current stage, which is counted for 5-6 years, will exclude the use of the fuel enriched by more than 20% from RF to other countries such as: Poland, Czeck Republick, Hungary, Roumania, Bulgaria, Libya, Viet-Nam, North Korea, Egypt, Latvia, Ukraine, Uzbekistan and Kazakhstan. In 1994 the Program, approved by RF Minatom authorities, has received the status of an inter-branch program since it was admitted by the RF Ministry for Science and Technical Policy. The Head of RF Minatom central administrative division N.I.Ermakov was nominated as the Head of the Russian Program, V.G.Aden, RDIPE Deputy Director, was nominated as the scientific leader. The Program was submitted to the Commission for Scientific, Technical and Economical Cooperation between USA and Russia headed by Vice-President A. Gore and Prime Minister V. Chemomyrdin and was given support also.

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

    Science.gov (United States)

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

    2015-09-01

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

  20. IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bostelmann, F. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on

  1. Spanish collaboration in the OECD Halden Reactor Project research on Gadolinia Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Horvath, M.; Munoz-Reja, C.; Tverberg, T.; Jenssen, H. K.

    2010-07-01

    Safe and reliable operation of nuclear power plants benefit from research and development advances and related technical solutions. One research platform is the OECD Halden Reactor Project (HRP). HRP is a joint undertaking of national organisations in 18 countries sponsoring a jointly financed programme under the auspices of the OECD - Nuclear Energy Agency (NEA). As a member state, Spain is participating HRP research programs with ENUSA as a partner in the fuel research programs. Improving the NPP operations, fuel cycles were designed to increase fuel burnup. Higher fuel burnup reduces the number of spent fuel assemblies and thus the costs of new fuel as well as the costs of back-end management. Higher burnup is reached either by prolonging the reactor cycles or by increasing the number of reactor cycles for the fuel in the core. Both ways entail additional requirements concerning fuel enrichment and burnable absorbers as additives and adjustments on the cladding material properties, such as mechanical treatment and chemical composition of the alloys. For these demands and needs ENUSA promotes the research on high burnup effects, gadolinium doped fuels and cladding material behaviour under irradiation. Various experiments, called IFA, are developed and performed also by providing materials. ENUSA collaborates with HRP on various experiments investigating the fuel densification and swelling, fission gas release, pressure limits on UO{sub 2} and (U,Gd)O{sub 2} fuels (IFA-504, -515, -636, -681); the cladding creep, lift-off, corrosion and hydrides on different tubing materials (IFA-567, -610, -638); instrumentation of the experiments, especially on pre-irradiated materials (IFA-533). These experiments are combined with model calculations to improve predictions for higher burnups and to maintain safety margins (IFA-515, -636, -681). Besides these unique in-pile experiments PIEs are performed as well on fuel and structural materials to complete the scope of these

  2. Novel Cryogenic Engineering Solutions for the New Australian Research Reactor Opal

    Science.gov (United States)

    Olsen, S. R.; Kennedy, S. J.; Kim, S.; Schulz, J. C.; Thiering, R.; Gilbert, E. P.; Lu, W.; James, M.; Robinson, R. A.

    2008-03-01

    In August 2006 the new 20MW low enriched uranium research reactor OPAL went critical. The reactor has 3 main functions, radio pharmaceutical production, silicon irradiation and as a neutron source. Commissioning on 7 neutron scattering instruments began in December 2006. Three of these instruments (Small Angle Neutron Scattering, Reflectometer and Time-of-flight Spectrometer) utilize cold neutrons. The OPAL Cold Neutron Source, located inside the reactor, is a 20L liquid deuterium moderated source operating at 20K, 330kPa with a nominal refrigeration capacity of 5 kW and a peak flux at 4.2meV (equivalent to a wavelength of 0.4nm). The Thermosiphon and Moderator Chamber are cooled by helium gas delivered at 19.8K using the Brayton cycle. The helium is compressed by two 250kW compressors (one with a variable frequency drive to lower power consumption). A 5 Tesla BSCCO (2223) horizontal field HTS magnet will be delivered in the 2nd half of 2007 for use on all the cold neutron instruments. The magnet is cooled by a pulse tube cryocooler operating at 20K. The magnet design allows for the neutron beam to pass both axially and transverse to the field. Samples will be mounted in a 4K to 800K Gifford-McMahon (GM) cryofurnace, with the ability to apply a variable electric field in-situ. The magnet is mounted onto a tilt stage. The sample can thus be studied under a wide variety of conditions. A cryogen free 7.4 Tesla Nb-Ti vertical field LTS magnet, commissioned in 2005 will be used on neutron diffraction experiments. It is cooled by a standard GM cryocooler operating at 4.2K. The sample is mounted in a 2nd GM cryocooler (4K-300K) and a variable electric field can be applied.

  3. Operational experiences and coping with ageing effects of the IRT-Sofia research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krezhov, K. [Inst. for Nuclear Research and Nuclear Energy, Sofia (Bulgaria)

    1995-12-31

    The present paper gives a review on the efforts to cope with reactor equipment ageing effects and describes the major experience gained in the maintenance work necessary to keep the reactor in good condition throughout the years. Also, a short description of the modernization project with preserved reactor power level of 2 MW is given. (orig.)

  4. Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

  5. Design of a Control Room for Jordan Research and Training Reactor (JRTR)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Jun; Suh, Sang Moon; Lee, Hyun Chul; Park, Je Yun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Since the main role of JRTR(Jordan Research and Training Reactor) operating personnel is safe and reliable operation of the reactor, MCR(Main Control Room) and SCR(Supplementary Control Room) must provide them with sufficient information and controls needed to optimize their performance. Before the TMI accident, control room were generally designed just with intuitive common sense, without using any proper HFE(human factors engineering) practices. Many results derived from the analysis of TMI accident showed that a more comprehensive and systematic approaches to develop MCR design requirements were needed. Moreover changes of operators' role as a decision maker from a physical controller in rapid improvement of control system which resulted in higher automation clearly needed more featured regulatory requirements and guidelines. So many regulatory and industrial guidance for control room design have been developed by relevant institution and regulatory bodies. In this paper, a conceptual design of the JRTR control room in the effort of satisfying current regulatory requirements and guidelines are presented. And some information display design is also presented

  6. The Multi一physics Research on I ron一Core Vibration Noise of Power Reactor

    Directory of Open Access Journals (Sweden)

    LI U Ja

    2017-02-01

    Full Text Available On the basis of theoretical research releted to the magnetostriction and maxwell’.s equations,the fi- nite element coupling in the transient electromagnetic field coupling,structure and sound field coupling has been developed In thts paper by using the flnlte element sOftWare CO}IS01., Whleh establish a serles three-phase COT’e re- actor model, to analyzing the power frequency magnetic field distribution,core magnetostrictive displacement,max- well force displacement and sound pressure level of the three-phase series core reactor under the power frequency working state. According to transient magnetic field distribution in the simulation of the reactor,the magnetic flux density distribution inside the reactor and the vibration displacement distribution are calculated,the acoustic field distribution is measured alao. It is shown that physical field simulation results and measured data are basically in consisent by experiment,it is proved multi-physics coupling is an effective method for forecast of noise.

  7. Intense positron source at the Munich research reactor FRM-II

    CERN Document Server

    Hugenschmidt, C; Schreckenbach, K; Strasser, B; Koegel, G; Sperr, P; Triftshaeuser, W

    2002-01-01

    The principle and the design of the in-pile positron source at the new Munich research reactor FRM-II are presented. Absorption of high-energy prompt gamma-rays from thermal neutron capture in sup 1 sup 1 sup 3 Cd generates positrons by pair production. For this purpose, a cadmium cap is placed inside the tip of the inclined beam tube SR11 in the neutron field of the reactor, where an undisturbed thermal neutron flux up to 2 x 10 sup 1 sup 4 n cm sup - sup 2 s sup - sup 1 is expected. At this position the flux ratio of thermal to fast neutrons will be better than 10 sup 4. Monte Carlo calculations showed that a mean capture rate in cadmium between 4.5 and 6.0 x 10 sup 1 sup 3 n cm sup - sup 2 s sup - sup 1 can be expected. Inside the cadmium cap a structure of platinum foils is placed for converting gamma-radiation into positron-electron pairs. The heated foils also act as positron moderators to generate monoenergetic positrons. After acceleration to 5 keV a positron beam is formed by electric lenses and guid...

  8. An overview of IPPE research on liquid metal fast reactor thermohydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Sorokin, A. P.; Efanov, A. D.; Zhukov, A. V.; Bogoslovskaia, G. P. [SSC RF-IPPE, Kaluga (Russian Federation)

    2003-07-01

    The paper presents brief information on the most significant researches in the fields of liquid metal hydrodynamics and heat transfer performed in the State Scientific Center of Russian Federation 'Institute for Physics and Power Engineering' named after A.I.Leypunski applied to sodium-cooled fast reactors. Experimental methods for studying liquid metal thermohydraulics and applied measurement techniques are overviewed briefly in the paper. Some results of fundamental thermohydraulic investigations, such as quasi-universal character of velocity and temperature profile in liquid metals, if considered normally to the channel wall etc. are presented. Specific features of heat transfer in liquid metal cooled fuel subassembly are mentioned, among them there are: high level of coolant temperature; significant influence of an interchannel exchange on velocity and temperature distribution; an availability of contact thermal resistance; large azimuthal non-uniformity of velocity and temperature; 'conjugate' problem of heat transfer in combined geometry of fuel pin; an absence of stabilization of heat transfer in non-standard channels; an influence of non-uniform heat generation. Special attention is given to the temperature fields in fuel subassembly subjected to deformation because of radioactive swelling and creeping, as well as in case of blockage of a part of subassembly cross section. Some results of thermohydraulic investigation are demonstrated for intermediate heat exchangers, pressurized head collectors. Also the developed methods and codes of thermohydraulic calculations applied to fast reactor core are considered: subchannel approach, porous body model.

  9. LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, W D; Goodman, R L; Heaberlin, S W; Hesson, G M; Nealley, C; Kirg, L L; Marshall, R K; McNair, G W; Meitzler, W D; Neally, G W; Parchen, L J; Pilger, J P; Rausch, W N; Russcher, G E; Schreiber, R E; Wildung, N J

    1981-02-01

    Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.

  10. 77 FR 68155 - The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R-84

    Science.gov (United States)

    2012-11-15

    ... COMMISSION The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R... Operating License No. R-84 (Application), which currently authorizes the Armed Forces Radiobiology Research... the renewal of Facility Operating License No. R-84, which currently authorizes the licensee to operate...

  11. Nuclear non-proliferation: the U.S. obligation to accept spent fuel from foreign research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shapar, Howard K.; Egan, Joseph R. [Shaw, Pittman, Potts and Trowbridge (United States)

    1995-12-31

    The U.S. Department of Energy (DOE) had a 35-year program for the sale and receipt (for reprocessing) of high-enriched research reactor fuel for foreign research reactors, executed pursuant to bilateral agreements with nuclear trading partners. In 1988, DOE abruptly let this program lapse, citing environmental obstacles. DOE promised to renew the program upon completion of an environmental review which was to take approximately six months. After three and a half years, an environmental assessment was finally produced.Over a year and half elapsed since publication of the assessment before DOE finally took action to renew the program. The paper sets forth the nuclear non-proliferation and related foreign policy considerations which support renewal of the program. It also summarized the contractual and other commitments made to foreign research reactors and foreign governments and aspects of U.S. environmental law as they apply to continuation of the program. (author).

  12. A binary mixed integer coded genetic algorithm for multi-objective optimization of nuclear research reactor fuel reloading

    Energy Technology Data Exchange (ETDEWEB)

    Binh, Do Quang [University of Technical Education Ho Chi Minh City (Viet Nam); Huy, Ngo Quang [University of Industry Ho Chi Minh City (Viet Nam); Hai, Nguyen Hoang [Centre for Research and Development of Radiation Technology, Ho Chi Minh City (Viet Nam)

    2014-12-15

    This paper presents a new approach based on a binary mixed integer coded genetic algorithm in conjunction with the weighted sum method for multi-objective optimization of fuel loading patterns for nuclear research reactors. The proposed genetic algorithm works with two types of chromosomes: binary and integer chromosomes, and consists of two types of genetic operators: one working on binary chromosomes and the other working on integer chromosomes. The algorithm automatically searches for the most suitable weighting factors of the weighting function and the optimal fuel loading patterns in the search process. Illustrative calculations are implemented for a research reactor type TRIGA MARK II loaded with the Russian VVR-M2 fuels. Results show that the proposed genetic algorithm can successfully search for both the best weighting factors and a set of approximate optimal loading patterns that maximize the effective multiplication factor and minimize the power peaking factor while satisfying operational and safety constraints for the research reactor.

  13. Assessment of the Implementation of a Neutron Measurement System During the Commissioning of the Jordan Research and Training Reactor

    Directory of Open Access Journals (Sweden)

    Sanghoon Bae

    2017-04-01

    Full Text Available The Jordan Research and Training Reactor (JRTR is the first research reactor in Jordan, the commissioning of which is ongoing. The reactor is a 5-MWth, open-pool type, light-water-moderated, and cooled reactor with a heavy water reflector system. The neutron measurement system (NMS applied to the JRTR employs a wide-range fission chamber that can cover from source range to power range. A high-sensitivity boron trifluoride counter was added to obtain more accurate measurements of the neutron signals and to calibrate the log power signals; the NMS has a major role in the entire commissioning stage. However, few case studies exist concerning the application of the NMS to a research reactor. This study introduces the features of the NMS and the boron trifluoride counter in the JRTR and shares valuable experiences from lessons learned from the system installation to its early commissioning. In particular, the background noise relative to the signal-to-noise ratio and the NMS signal interlock are elaborated. The results of the count rates with the neutron source and the effects of the discriminator threshold are summarized.

  14. Replacement Nuclear Research Reactor. Supplement to Draft Environmental Impact Statement. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-01-01

    The Draft Environmental Impact Statement for a replacement research reactor at Lucas Heights, was available for public examination and comment for some three months during 1998. A Supplement to the Draft Environmental Impact Statement (Draft EIS) has been completed and was lodged with Environment Australia on 18 January 1999. The Supplement is an important step in the overall environmental assessment process. It reviews submissions received and provides the proponent`s response to issues raised in the public review period. General issues extracted from submissions and addressed in the Supplement include concern over liability issues, Chernobyl type accidents, the ozone layer and health issues. Further studies, relating to issues raised in the public submission process, were undertaken for the Supplementary EIS. These studies confirm, in ANSTO`s view, the findings of the Draft EIS and hence the findings of the Final EIS are unchanged from the Draft EIS

  15. Fuel management optimization for the WWR-M research reactor in Kiev

    Energy Technology Data Exchange (ETDEWEB)

    Mahlers, Y.P. [Institute for Nuclear Research Prospect Nauki 47, Kiev 252022 (Ukraine)

    2002-07-01

    Core loading patterns, number and types of fuel assemblies in the core as well as discharged fuel burnup are determined for the WWR-M research reactor in Kiev by the optimization procedure providing high neutron flux under the safety and fuel constraints. For neutronics calculation, the iterational hybrid method combining diffusion model with higher approximations of neutron transport equation is applied. The results of calculation are shown to be consistent with the results of measurement. To determine the best placement of fuel assemblies in the core, successive mixed-integer linear programming and backward diffusion calculation is used. An example of maximization of thermal neutron flux in large channels in the core is demonstrated. (author)

  16. A new small-angle neutron scattering spectrometer at China Mianyang research reactor

    Science.gov (United States)

    Peng, Mei; Sun, Liangwei; Chen, Liang; Sun, Guangai; Chen, Bo; Xie, Chaomei; Xia, Qingzhong; Yan, Guanyun; Tian, Qiang; Huang, Chaoqiang; Pang, Beibei; Zhang, Ying; Wang, Yun; Liu, Yaoguang; Kang, Wu; Gong, Jian

    2016-02-01

    A new pinhole small-angle neutron scattering (SANS) spectrometer, installed at the cold neutron source of the 20 MW China Mianyang Research Reactor (CMRR) in the Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, has been put into use since 2014. The spectrometer is equipped with a multi-blade mechanical velocity selector, a multi-beam collimation system, and a two-dimensional He-3 position sensitive neutron detector. The q-range of the spectrometer covers from 0.01 nm-1 to 5.0 nm-1. In this paper, the design and characteristics of the SANS spectrometer are described. The q-resolution calculations, together with calibration measurements of silver behenate and a dispersion of nearly monodisperse poly-methyl-methacrylate nanoparticles indicate that our SANS spectrometer has a good performance and is now in routine service.

  17. Available reprocessing and recycling services for research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tozser, Sandor; Marshall, Frances M.; Adelfang, Pablo; Bradley, Edward [International Atomic Energy Agency, Vienna (Austria); Budu, Madalina Elena [SOSNY Research and Development Company, Moscow (Russian Federation); Chiguer, Mustapha [AREVA, Paris La Defense (France)

    2016-03-15

    International activities in the back end of the research reactor (RR) fuel cycle have so far been dominated by the programmes of acceptance of highly-enriched uranium (HEU) spent nuclear fuel (SNF) by the country where it was originally enriched. In the future inventories of LEU SNF will continue to be created and the back end solution of RR SNF remains a critical issue. The IAEA, based on the experience gained during the decade of international cooperation in supporting the objectives of the HEU take-back programmes, drew up a report presenting available reprocessing and recycling services for RR SNF. This paper gives an overview of the report, which will address all aspects of reprocessing and recycling services for RR SNF.

  18. Elements of record management system for the RA research reactor decommissioning

    Directory of Open Access Journals (Sweden)

    Stejić Milijana

    2004-01-01

    Full Text Available According to latest recommendations, the record management system of a nuclear facility should operate as a part of the integrated management information system, and is implemented at the very beginning of the facility’s life cycle. The record management becomes particularly important at the end of the operation of a facility and then the operational record management system gradually transforms to a decommissioning one. However there is a significant number of nuclear facilities in the world which have reached the decommissioning stage with out having neither the initial decommissioning plan nor the established record management system. The objective of this paper is to introduce constituted elements of the record management system for the decommissioning of the RA research reactor in the VINČA Institute of Nuclear Sciences, and to discuss future planned actions related to this matter.

  19. 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.

  20. Generation IV Reactor Safety and Materials Research by the Institute for Energy and Transport at the European Commission's Joint Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Tuček, K., E-mail: kamil.tucek@ec.europa.eu; Tsige-Tamirat, H.; Ammirabile, L.; Lázaro, A.; Grah, A.; Carlsson, J.; Döderlein, Ch.; Oettingen, M.; Fütterer, M.A.; D’Agata, E.; Laurie, M.; Turba, K.; Ohms, C.; Nilsson, K.-F.; Hähner, P.

    2013-12-15

    To support the drafting, development, implementation and monitoring of European energy and transport policy, the Institute for Energy and Transport of the European Commissions’ Joint Research Centre conducts pre-competitive research in the areas of experimental qualification of advanced fuels and materials as well as simulation and modelling of reactor safety and material performance. The work covers assessments, design optimisation and improvements to the safety and performance of new, innovative reactor systems, materials and instrumentation, in order to meet the EU's long-term energy needs while respecting enhanced safety, sustainability, and economic aspects. The research is linked, and contributes, to related EURATOM Framework Programme projects, Generation IV International Forum (GIF), International Atomic Energy Agency as well as OECD's Nuclear Energy Agency (OECD NEA) activities. The current paper gives an overview and examples of past, current, and upcoming activities in the areas of reactor safety assessments, advanced fuel irradiation and materials research.

  1. Very High Temperature Reactor (VHTR) Survey of Materials Research and Development Needs to Support Early Deployment

    Energy Technology Data Exchange (ETDEWEB)

    Eric Shaber; G. Baccaglini; S. Ball; T. Burchell; B. Corwin; T. Fewell; M. Labar; P. MacDonald; P. Rittenhouse; Russ Vollam; F. Southworth

    2003-01-01

    The VHTR reference concept is a helium-cooled, graphite moderated, thermal neutron spectrum reactor with an outlet temperature of 1000 C or higher. It is expected that the VHTR will be purchased in the future as either an electricity producing plant with a direct cycle gas turbine or a hydrogen producing (or other process heat application) plant. The process heat version of the VHTR will require that an intermediate heat exchanger (IHX) and primary gas circulator be located in an adjoining power conversion vessel. A third VHTR mission - actinide burning - can be accomplished with either the hydrogen-production or gas turbine designs. The first ''demonstration'' VHTR will produce both electricity and hydrogen using the IHX to transfer the heat to either a hydrogen production plant or the gas turbine. The plant size, reactor thermal power, and core configuration will be designed to assure passive decay heat removal without fuel damage during accidents. The fuel cycle will be a once-through very high burnup low-enriched uranium fuel cycle. The purpose of this report is to identify the materials research and development needs for the VHTR. To do this, we focused on the plant design described in Section 2, which is similar to the GT-MHR plant design (850 C core outlet temperature). For system or component designs that present significant material challenges (or far greater expense) there may be some viable design alternatives or options that can reduce development needs or allow use of available (cheaper) materials. Nevertheless, we were not able to assess those alternatives in the time allotted for this report and, to move forward with this material research and development assessment, the authors of this report felt that it was necessary to use a GT-MHR type design as the baseline design.

  2. Euratom research and training in nuclear reactor safety: Towards European research and the higher education area

    Energy Technology Data Exchange (ETDEWEB)

    Goethem, G. van [Nuclear Fission and Radiation Protection European Commission, Building MO75-5-34, B-1049 Brussels (Belgium)]. E-mail: georges.van-goethem@cec.eu.int

    2004-07-01

    In this invited lecture, research and training in nuclear fission are looked at from a European perspective with emphasis on the three success factors of any European policy, namely: common needs, vision and instruments, that ought to be strongly shared amongst the stakeholders across the Member States concerned. As a result, the following questions are addressed: What is driving the current EU trend towards more research, more education and more training, in general? Regarding nuclear fission, in particular, who are the end-users of Euratom 'research and training' and what are their expectations from EU programmes? Do all stakeholders share the same vision about European research and training in nuclear fission? What are the instruments proposed by the European Commission (EC) to conduct joint research programmes of common interest for the nuclear fission community? In conclusion, amongst the stakeholders in Europe, there seems to be a wide consensus about common needs and instruments, but not about a common vision regarding nuclear. (author)

  3. RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 3: Verification and Validation Report.

    Energy Technology Data Exchange (ETDEWEB)

    Talley, Darren G.

    2017-04-01

    This report describes the work and results of the verification and validation (V&V) of the version 1.0 release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code shows good agreement between simulation and actual ACRR operations.

  4. Design and development of fast pneumatic transfer system (PTS) for instrumental neutron activation analysis at Jordan research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Yongsam; Kim, Sunha; Moon, Jonghwa; Choi, Jinbok; Lee, Jongmin; Ryu, Jungsu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    A pneumatic transfer system (PTS) is one of the important equipment used for an neutron irradiation of a target material for an instrumental neutron activation analysis (INAA) in a research reactor. In particular, a rapid pneumatic transportation of irradiation capsule is essential for an accurate measurement of a short half-life nuclide. Three types of PTS for NAA facility at the Jordan Research and Training Reactor (JRTR) were newly developed for a functional improvement involving a manual and an automatic system which is equipped with programmable logic controller, software, and 13 devices to facilitate optimal operation of the system. In this paper, the designs and construction of these PTS, the operation and control of the system are described. In addition, a functional and operational test of the system were carried out as one of the basic requirement and characteristic parameters, and the results were reported to provide a user information as well as for the management and safety of the reactor.

  5. Development of technology for next generation reactor - Research of evaluation technology for nuclear power plant -

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)] [and others

    1993-09-01

    For development of next generation reactor, a project for evaluation technology for nuclear power plant is performed. Evaluation technology is essential to next generation reactor for reactor safety and system analysis. For design concept, detailed evaluation technologies are studied as follows: evaluation of safety margin, evaluation of safety facilities, evaluation of measurement and control technology; man-machine interface. Especially for thermal efficiency, thermal properties and chemical composition of inconel 690 tube, instead of inconel 600 tube, are measured for steam generator. (Author).

  6. Training courses on neutron detection systems on the ISIS research reactor: on-site and through internet training

    Energy Technology Data Exchange (ETDEWEB)

    Lescop, B.; Badeau, G.; Ivanovic, S.; Foulon, F. [National Institute for Nuclear science and Technology French Atomic Energy and Alternative Energies Commission (CEA), Saclay Research Center, 91191 Gif-sur-Yvette (France)

    2015-07-01

    Today, ISIS research reactor is an essential tool for Education and Training programs organized by the National Institute for Nuclear Science and Technology (INSTN) from CEA. In the field of nuclear instrumentation, the INSTN offers both, theoretical courses and training courses on the use of neutron detection systems taking advantage of the ISIS research reactor for the supply of a wide range of neutron fluxes. This paper describes the content of the training carried out on the use of neutron detectors and detection systems, on-site or remote. The ISIS reactor is a 700 kW open core pool type reactor. The facility is very flexible since neutron detectors can be inserted into the core or its vicinity, and be used at different levels of power according to the needs of the course. Neutron fluxes, typically ranging from 1 to 10{sup 12} n/cm{sup 2}.s, can be obtained for the characterisation of the neutron detectors and detection systems. For the monitoring of the neutron density at low level of power, the Instrumentation and Control (I and C) system of the reactor is equipped with two detection systems, named BN1 and BN2. Each way contains a fission chamber, type CFUL01, connected to an electronic system type SIREX.The system works in pulse mode and exhibits two outputs: the counting rate and the doubling time. For the high level of power, the I and C is equipped with two detection systems HN1 and HN2.Each way contain a boron ionization chamber (type CC52) connected to an electronics system type SIREX. The system works in current mode and has two outputs: the current and the doubling time. For each mode, the trainees can observe and measure the signal at the different stages of the electronic system, with an oscilloscope. They can understand the role of each component of the detection system: detector, cable and each electronic block. The limitation of the detection modes and their operating range can be established from the measured signal. The trainees can also

  7. Nuclear Research Reactor IEA-R1 - A Study of the Preparing for Decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Lopes, Valdir Maciel; Filho, Tufic Madi; Ricci, Walter; Martins, Mauro Onofre; Pereira, Maria da Conceicao [Instituto de Pesquisas Energeticas e Nucleares IPEN/CNEN - Sao Paulo, Av. Professor Lineu Prestes n 2242, USP, Sao Paulo (Brazil)

    2015-07-01

    The Institute of Energy and Nuclear Research (IPEN), Sao Paulo, according to the assignments given by the National Commission of Nuclear Energy (CNEN), enabled the development of this study, especially operational reports about refurbishing carried out on 2013, involving the production of radioisotopes and research in the areas of Radiochemistry and Nuclear Physics. These reports are made in accordance with established standard procedures to meet the requirements of CNEN (National Nuclear Energy Commission, the regulator the nuclear area activities in Brazil) and IAEA (International Atomic Energy Agency). This study presents an assessment of the procedures and methods of treatments for decontamination of the refrigeration primary circuit and changes parts, equipment and tubes of the of the IEA-R1 nuclear research reactor, pool type, power between 3,5 and 4,5 MW. In order to have a sequence in the work, the well-known contaminant radioisotopes were evaluated firstly, using Geiger- Muller equipment. In the second phase, the decontamination was done manually together with the ultrasound cleaning and washing equipment. From the several water solutions of citric acid assessment, the concentration with better confidence was obtained; in order to achieve the best results for decontamination. This study intends to define the best process for decontamination with low taxes of waste and without expensive costs. (authors)

  8. Interim status report on lead-cooled fast reactor (LFR) research and development.

    Energy Technology Data Exchange (ETDEWEB)

    Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.; Smith, C. F.; de Caro, M.; Halsey, W. G.; Li, N.; Hosemann, P.; Zhang, J.; Bolind, A.; LLNL; LANL; Univ. of Illinois

    2008-03-31

    This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigation of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup

  9. Annual report on reactor safety research projects. Reporting period 2014. Progress report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    Within its competence for energy research the Federal Ministry for Economic Affairs and Energy (BMWi) sponsors research projects on the safety of nuclear power plants currently in operation. The objective of these projects is to provide fundamental knowledge, procedures and methods to contribute to realistic safety assessments of nuclear installations, to the further development of safety technology and to make use of the potential of innovative safety-related approaches. The Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, by order of the BMWi, continuously issues information on the status of such research projects by publishing semi-annual and annual progress reports within the series of GRS-F-Fortschrittsberichte (GRS-F-Progress Reports). Each progress report represents a compilation of individual reports about the objectives, work performed, results achieved, next steps of the work etc. The individual reports are prepared in a standard form by the research organisations themselves as documentation of their progress in work. The progress reports are published by the Project Management Agency/Authority Support Division of GRS. The reports as of the year 2000 are available in the lnternet-based information system on results and data of reactor safety research (http://www.grs-fbw.de). The compilation of the reports is classified according to the classification system ''Joint Safety Research Index (JSRI)''. The reports are arranged in sequence of their project numbers. lt has to be pointed out that the authors of the reports are responsible for the contents of this compilation. The BMWi does not take any responsibility for the correctness, exactness and completeness of the information nor for the observance of private claims of third parties.

  10. Euratom innovation in nuclear fission: Community research in reactor systems and fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Goethem, G. van [European Commission, DG Research, Directorate J: Energy (Euratom), 1049 Brussels (Belgium)]. E-mail: georges.van-goethem@ec.europa.eu; Hugon, M. [European Commission, DG Research, Directorate J: Energy (Euratom), 1049 Brussels (Belgium); Bhatnagar, V. [European Commission, DG Research, Directorate J: Energy (Euratom), 1049 Brussels (Belgium); Manolatos, P. [European Commission, DG Research, Directorate J: Energy (Euratom), 1049 Brussels (Belgium); Deffrennes, M. [European Commission, DG Research, Directorate J: Energy (Euratom), 1049 Brussels (Belgium)

    2007-07-15

    The following questions are naturally at the heart of the current Euratom research and training framework programme:(1)What are the challenges facing the European Union nuclear fission research community in the short (today), medium (2010) and long term (2040)? (2)What kind of research and technological development (RTD) does Euratom offer to respond to these challenges, in particular in the area of reactor systems and fuel cycles? In the general debate about energy supply technologies there are challenges of both a scientific and technological (S/T) as well as an economic and political (E/P) nature. Though the Community research programme acts mainly on the former, there is nevertheless important links with Community policy. These not only exist in the specific area of nuclear policy, but also more generally as is depicted in the following figure. It is shown in the particular area of nuclear fission, to what extent Euratom research, education and innovation ('Knowledge Triangle' in above figure) respond to the following long-term criteria: (1) sustainability, (2) economics, (3) safety, and (4) proliferation resistance. Research and innovation in nuclear fission technology has broad and extended geographical, disciplinary and time horizons:- the community involved extends to all 25 EU Member States and beyond; - the research assembles a large variety of scientific disciplines; - three generations of nuclear power technologies (called II, III and IV) are involved, with the timescales extending from now to around the year 2040. To each of these three generations, a couple of challenges are associated (six in total):- Generation II (1970-2000, today): security of supply+environmental compatibility; - Generation III (around 2010): enhanced safety and competitiveness (economics); - Generation IV (around 2040): cogeneration of heat and power, and full recycling. At the European Commission (EC), the research related to nuclear reactor systems and fuel cycles is

  11. Development of Thermal Stress Test Profile for Class 1E Equipment in Main Control Room of a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Seung Ki; Park, Je Yun; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jang, Hee Jun [Daewoo Engineering and Construction Co, Seoul (Korea, Republic of)

    2014-10-15

    The primary objective of an equipment qualification (EQ) is to demonstrate with reasonable assurance that Class 1E equipment for which a qualified life or condition has been established can perform its safety function without experiencing common-cause failures before, during, and after applicable design basis events (DBEs). For the environmental test of Class 1E equipment installed in the main control room (MCR) of a research reactor, the thermal stress test profile should be developed based on normal and abnormal environmental conditions of the MCR. A thermal stress test profile for Class 1E equipment located in the MCR of a research reactor is developed based on an engineering justification. A set of thermal stress test conditions including temperature, test duration, and number of test cycles is defined considering the specific characteristics of the targeted research reactor. The method used in this study can be applied to develop a thermal stress test profile used for the EQ test to qualify Class 1E equipment installed in the MCR of various research reactors.

  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... renewal of Facility Operating License No. R-108 (``Application''), which currently authorizes the Dow... Operating License No. R-108 for the DTRR. The application contains SUNSI. Based on its initial review of the...

  13. Perspective on fusion research in China (2) fusion activities in China with special intonation on hybrid reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Lijian, Qiu

    2001-09-01

    Chinese fusion research was started from 1958, but with more clear problem definition it has been set up as the national program for development of the hybrid reactor in 1986. In this paper, it will be described how the organized program is going on.

  14. Awareness, Preference, Utilization, and Messaging Research for the Spallation Neutron Source and High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bryant, Rebecca [Bryant Research, LLC; Kszos, Lynn A [ORNL

    2011-03-01

    Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one

  15. Conceptual Thermal Hydraulic Design of a 20MW Multipurpose Research Reactor (KAERI/VAEC joint study on a new research reactor for Vietnam)

    Energy Technology Data Exchange (ETDEWEB)

    Chae, Hee Taek; Seo, Chul Gyo; Park, Jong Hark; Park, Cheol [Kaeri, Daejeon (Korea, Republic of); Vinh, Le Vinh; Nghiem, Huynh Ton; Dang, Vo Doan Hai [Dalat Nuclear Research Reactor, Hanoi (Viet Nam)

    2007-08-15

    The conceptual thermal hydraulics design analyses for the 20 MW reference AHR core have been jointly performed by the KAERI and DNRI(VAEC). The preliminary core thermal hydraulic characteristics and safety margins for the AHR core were studied for various core flow rates, fuel assembly powers and core inlet temperatures. Statistical method was applied to the thermal hydraulic design of the reactor core. The MATRA{sub h} subchannel code has been applied to evaluate the thermal hydraulic performances of the AHR and the resulting thermal margins of the core under the forced convection cooling mode during a nominal power operation and the natural circulation mode during a reactor shutdown condition. In addition, typical accident analyses were carried out for a loss of flow accident by a primary pump seizure and a reactivity induced accident by a CAR rod withdrawal during a normal full power operation. The normal full power operation of the AHR was ensured with a sufficient safety margin for the onset of nucleate boiling phenomena. The AHR also had a sufficient natural circulation cooling capability to cool the core without the onset of nucleate boiling in the channel after a normal reactor shutdown and the anticipated transients. It was confirmed by the typical accident analyses that the AHR core was sufficiently protected from the loss of flow by the primary cooling pump seizure and the overpower transients by the CAR withdrawal from the MCHFR and fuel temperature points of view.

  16. International Atomic Energy Agency (IAEA) Coordinated Research Projects on Structural Integrity of Reactor Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Server, W. L. [ATI Consulting, Pinehurst, NC; Nanstad, Randy K [ORNL

    2009-01-01

    The International Atomic Energy Agency (IAEA) has conducted a series of Coordinated Research Projects (CRPs) that have focused on irradiated reactor pressure vessel (RPV) steel fracture toughness properties and approaches for assuring structural integrity of RPVs throughout operating life. A series of nine CRPs have been sponsored by the IAEA, starting in the early 1970s, focused on neutron radiation effects on RPV steels. The purpose of the CRPs was to develop comparisons and correlations to test the uniformity of irradiated results through coordinated international research studies and data sharing. Consideration of dose rate effects, effects of alloying (nickel, manganese, silicon, etc.) and residual elements (eg., copper and phosphorus), and drop in upper shelf toughness are also important for assessing neutron embrittlement effects. The ultimate use of embrittlement understanding is assuring structural integrity of the RPV under current and future operation and accident conditions. Material fracture toughness is the key ingredient needed for this assessment, and many of the CRPs have focused on measurement and application of irradiated fracture toughness. This paper presents an overview of the progress made since the inception of the CRPs in the early 1970s. The chronology and importance of each CRP have been reviewed and put into context for continued and long-term safe operation of RPVs.

  17. Development of Neutron Imaging System for Neutron Tomography at Thai Research Reactor TRR-1/M1

    Science.gov (United States)

    Wonglee, S.; Khaweerat, S.; Channuie, J.; Picha, R.; Liamsuwan, T.; Ratanatongchai, W.

    2017-09-01

    The neutron imaging is a powerful non-destructive technique to investigate the internal structure and provides the information which is different from the conventional X-ray/Gamma radiography. By reconstruction of the obtained 2-dimentional (2D) images from the taken different angle around the specimen, the tomographic image can be obtained and it can provide the information in more detail. The neutron imaging system at Thai Research Reactor TRR-1/M1 of Thailand Institute of Nuclear Technology (Public Organization) has been developed to conduct the neutron tomography since 2014. The primary goal of this work is to serve the investigation of archeological samples, however, this technique can also be applied to various fields, such as investigation of industrial specimen and others. This research paper presents the performance study of a compact neutron camera manufactured by Neutron Optics such as speed and sensitivity. Furthermore, the 3-dimentional (3D) neutron image was successfully reconstructed at the developed neutron imaging system of TRR-1/M1.

  18. Regulatory Approach to Safety of Long Time Operating Research Reactors in Russia

    Energy Technology Data Exchange (ETDEWEB)

    Sapozhnikov, Alexander [Industrial and Nuclear Supervision Service, Moscow (Russian Federation)

    2013-07-01

    In the Russian Federation more than 60% of operating Nuclear Research Facilities (NRFs) are of age over 30 years old or their usage exceeds originally conceived continuous operation. In this regard, important areas of regulatory body activity are: 1) a systematic assessment of the actual state of structures, systems and components (SSCs) important to safety, 2) control of implementation of organizational and technical measures to mitigate ageing impact on the basis of programmes to manage reliability (service life) of SSCs, and 3) issues of facility modification/reconstruction in line with up-to-day safety requirements. The practice of licensing NRFs with long operating times shows that the national regulations are generally in compliance with IAEA recommendations for ageing management of research reactors. In operating organizations, the ageing management is being effectively provided as a part of the integrated management system for NRFs, including the monitoring of the reliability of SSCs, a methodology to detect their ageing, reporting and investigation of events, analysis of their root causes, and measures to prevent and mitigate ageing effects to safety. The report outlines a good practice of safety regulation of NRFs with long operating times and based on lessons learned from experience, including challenges for future improvement of ageing management.

  19. Study on Pressure drop for Ion Exchanger in Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ki-jung; Choi, Jungwoon; Kim, Seong-Hoon; Chi, Dae-Young; Park, Cheol [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The Jordan Research and Training Reactor (JRTR) is currently being constructed and commissioned in the JUST (Jordan University of Science and Technology) site. The main fluid systems relevant to the JRTR have been proceeding at the Korea Atomic Energy Research Institute. In order to achieve the purpose of the pool water purification, two filters and two ion exchangers which can be to remove suspended solids and ionic impurities in the in-taken pool water have been designed. For the reliable design of this system pump, it is important to predict the pressure drop of the system equipment including the ion exchanger. In this study, the pressure drop in the ion exchanger of PWMS is predicted by using the well-known model and the results provided from manufacturing company. And, the calculated results are compared to the actual data which is measured from the ion exchanger during the PWMS commissioning. The predicted pressure drop is dominated by the resin bed as a portion of about 85% for total pressure drop. The predicted pressure drop is compared to the measured pressure drop of the ion exchanger which is installed in the JRTR, the data above 5 kg/s agree within 5% in the entire range.

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

    Science.gov (United States)

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

    2016-10-01

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

  1. Neutronic design study of accelerator driven system (ADS) for Jordan subcritical reactor as a neutron source for nuclear research.

    Science.gov (United States)

    Xoubi, Ned

    2018-01-01

    In this paper, a preliminary neutronic design study of an accelerator driven subcritical system for Jordan Subcritical Assembly (JSA) is presented. The conceptual design of coupling the JSA core with proton accelerator and spallation target is investigated, and its feasibility as a neutron source for nuclear research, and possibly for target irradiation and isotope production evaluated. 3D MCNPX model of the JSA reactor, the accelerator beam, and the Pb target was developed, based on actual reactor parameters. MCNPX calculations were carried out to estimate the absolute radial and axial neutron flux in the reactor, and to calculate the multiplication factor K eff and heat generated in the reactor. Numerical results showed an enormous increase in the neutron flux, by seven orders of magnitude, compared to the current JSA core design using Pu-Be source. In this research the results obtained are discussed and compared with those of the JSA, and do confirm the feasibility of utilizing the JSA as a viable nuclear research facility with adequate neutron flux. Copyright © 2017 Elsevier Ltd. All rights reserved.

  2. An Integrated Management System (IMS) for JM-1 SLOWPOKE-2 research reactor in Jamaica: experiences in documentation

    Energy Technology Data Exchange (ETDEWEB)

    Warner, T., E-mail: traceyann.warner02@uwimona.edu.jm [Univ. of West Indies, Mona (Jamaica)

    2014-07-01

    Since the first criticality in March 1984, the Jamaica SLOWPOKE-2 research reactor at the University of the West Indies, Mona located in the department of the International Centre for Environmental and Nuclear Sciences (ICENS) has operated for approximately 52% of the lifetime of the existing core configuration. The 20kW pool type research reactor has been primarily used for neutron activation analysis in environmental, agricultural, geochemical, health-related studies and mineral exploration in Jamaica. The involvement of the JM-1 reactor for research and teaching activities has segued into commercial applications which, coupled with the current core conversion programme from HEU to LEU, has demanded the implementation of management systems to satisfy regulatory requirements and assure compliance with internationally defined quality standards. At ICENS, documentation related to the Quality Management System aspect of an Integrated Management System (IMS) is well underway. The quality system will incorporate operational and nuclear safety, training, maintenance, design, utilization, occupational health and safety, quality service, and environmental management for its Nuclear Analytical Laboratory, NAL. The IMS is being designed to meet the requirements of the IAEA GS-R-3 with additional controls from international standards including: ISO/IEC 17025:2005, ISO 9001:2008, ISO 14001:2004 and OHSAS 18001:2007. This paper reports on the experiences of the documentation process in a low power reactor facility characterized by limited human resource, where innovative mechanisms of system automation and modeling are included to increase productivity and efficiency. (author)

  3. New reactor concepts. An analysis of the actual research status; Neue Reaktorkonzepte. Eine Analyse des aktuellen Forschungsstands

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph; Englert, Matthias

    2017-04-15

    The report on new reactor concepts covers the following issues: characterization and survey of new reactor concepts; evaluation criteria: safety, resources for fuel supply, waste problems, economy and proliferation; comprehensive relevant aspects: thorium as alternative resource, partitioning and transmutation; actual developments and preliminary experiences for fast breeding reactor (FBR), high-temperature reactor (HTR), molten salt reactor (MSR), small modular reactor (SMR).

  4. Design and R&D Progress of China Lead-Based Reactor for ADS Research Facility

    Directory of Open Access Journals (Sweden)

    Yican Wu

    2016-03-01

    Full Text Available In 2011, the Chinese Academy of Sciences launched an engineering project to develop an accelerator-driven subcritical system (ADS for nuclear waste transmutation. The China Lead-based Reactor (CLEAR, proposed by the Institute of Nuclear Energy Safety Technology, was selected as the reference reactor for ADS development, as well as for the technology development of the Generation IV lead-cooled fast reactor. The conceptual design of CLEAR-I with 10 MW thermal power has been completed. KYLIN series lead-bismuth eutectic experimental loops have been constructed to investigate the technologies of the coolant, key components, structural materials, fuel assembly, operation, and control. In order to validate and test the key components and integrated operating technology of the lead-based reactor, the lead alloy-cooled non-nuclear reactor CLEAR-S, the lead-based zero-power nuclear reactor CLEAR-0, and the lead-based virtual reactor CLEAR-V are under realization.

  5. Decommissioning of the ASTRA research reactor: Dismantling the auxiliary systems and clearance and reuse of the buildings

    Directory of Open Access Journals (Sweden)

    Meyer Franz

    2008-01-01

    Full Text Available The paper presents work performed in the last phase of the decommissioning of the ASTRA research reactor at the Austrian Research Centers Seibersdorf. Dismantling the pump room installations and the ventilation system, as well as the clearance of the buildings is described. Some conclusions and summary data regarding the timetable, material management, and the cost of the entire project are also presented.

  6. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, January--March 1976

    Energy Technology Data Exchange (ETDEWEB)

    Zane, J. O.; Farman, R. F.; Hanson, D. J.; Peterson, A. C.; Ybarrondo, L. J.; Berta, V. T.; Naff, S. A.; Crocker, J. G.; Martinson, Z. R.; Smolik, G. R.; Cawood, G. W.; Quapp, W. J.; Ramsthaler, J. H.; Ransom, V. H.; Scofield, M. P.; Dearien, J. A.; Bohn, M. P.; Burnham, B. W.; James, S. W.; Lee, W. H.; Lime, J. F.; Nalezny, C. L.; MacDonald, P. E.; Thompson, L. B.; Domenico, W. F.; Rice, R. E.; Hendrix, C. E.; Davis, C. B.

    1976-06-01

    Light water reactor sfaety research performed January through March 1976 is summarized. Results of the Semiscale Mod-1 blowdown heat transfer test series relating to those phenomena that influence core fluid and heat transfer effects are analyzed, and preliminary analyses of the recently completed reflood heat transfer test series are summarized for the forced and gravity feed reflood tests. The first nonnuclear LOCE in the LOFT program was successfully completed and preliminary results are presented. Preliminary results are given for the PCM 8-1 RF Test, the PCM-2A Test, and the Irradiation Effects Scoping Test 2 in the Thermal Fuel Behavior Program. Model development and verification efforts reported in the Reactor Behavior Program include checkout of RELAP4/MOD5 Update 1, development of a new hydrodynamic model for two-phase separated flows, development of the RACHET code to assess the assumptions in current fuel behavior codes of uniform stress and strain in the cladding, modifications of the containment code BEACON, analysis of results from the Halden Assembly IFA-429 helium sorption experiment, development of correlations for the thermal conductivity of UO/sub 2/ and (U,Pu)O/sub 2/, and evaluation of RALAP4 through comparison of calculated results with data from the GE Blowdown Heat Transfer and Semiscale experiments.

  7. State of the art on nuclear heating measurement methods and expected improvements in zero power research reactors

    Directory of Open Access Journals (Sweden)

    Le Guillou Mael

    2017-01-01

    Full Text Available The paper focuses on the recent methodological advances suitable for nuclear heating measurements in zero power research reactors. This bibliographical work is part of an experimental approach currently in progress at CEA Cadarache, aiming at optimizing photon heating measurements in low-power research reactors. It provides an overview of the application fields of the most widely used detectors, namely thermoluminescent dosimeters (TLDs and optically stimulated luminescent dosimeters. Starting from the methodology currently implemented at CEA, the expected improvements relate to the experimental determination of the neutron component, which is a key point conditioning the accuracy of photon heating measurements in mixed n–γ field. A recently developed methodology based on the use of 7Li and 6Li-enriched TLDs, precalibrated both in photon and neutron fields, is a promising approach to deconvolute the two components of nuclear heating. We also investigate the different methods of optical fiber dosimetry, with a view to assess the feasibility of online photon heating measurements, whose primary benefit is to overcome constraints related to the withdrawal of dosimeters from the reactor immediately after irradiation. Moreover, a fibered setup could allow measuring the instantaneous dose rate during irradiation, as well as the delayed photon dose after reactor shutdown. Some insights from potential further developments are given. Obviously, any improvement of the technique has to lead to a measurement uncertainty at least equal to that of the currently used methodology (∼5% at 1σ.

  8. Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Baek, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, L. Y. [Brookhaven National Lab. (BNL), Upton, NY (United States); Diamond, D. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-08-30

    An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enriched uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.

  9. Research and Development Methodology for Practical Use of Accident Tolerant Fuel in Light Water Reactors

    Directory of Open Access Journals (Sweden)

    Masaki Kurata

    2016-02-01

    Full Text Available Research and development (R&D methodology for the practical use of accident tolerant fuel (ATF in commercial light water reactors is discussed in the present review. The identification and quantification of the R&D-metrics and the attribute of candidate ATF-concepts, recognition of the gap between the present R&D status and the targeted practical use, prioritization of the R&D, and technology screening schemes are important for achieving a common understanding on technology screening process among stakeholders in the near term and in developing an efficient R&D track toward practical use. Technology readiness levels and attribute guides are considered to be proper indices for these evaluations. In the midterm, the selected ATF-concepts will be developed toward the technology readiness level-5, at which stage the performance of the prototype fuel rods and the practicality of industrial scale fuel manufacturing will be verified and validated. Regarding the screened-out concepts, which are recognized to have attractive potentials, the fundamental R&D should be continued in the midterm to find ways of addressing showstoppers.

  10. Development of a computational database for probabilistic safety assessment of nuclear research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macedo, Vagner S.; Oliveira, Patricia S. Pagetti de; Andrade, Delvonei Alves de, E-mail: vagner.macedo@usp.br, E-mail: patricia@ipen.br, E-mail: delvonei@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The objective of this work is to describe the database being developed at IPEN - CNEN / SP for application in the Probabilistic Safety Assessment of nuclear research reactors. The database can be accessed by means of a computational program installed in the corporate computer network, named IPEN Intranet, and this access will be allowed only to professionals previously registered. Data updating, editing and searching tasks will be controlled by a system administrator according to IPEN Intranet security rules. The logical model and the physical structure of the database can be represented by an Entity Relationship Model, which is based on the operational routines performed by IPEN - CNEN / SP users. The web application designed for the management of the database is named PSADB. It is being developed with MySQL database software and PHP programming language is being used. Data stored in this database are divided into modules that refer to technical specifications, operating history, maintenance history and failure events associated with the main components of the nuclear facilities. (author)

  11. INTERIM STORAGE AND LONG TERM DISPOSAL OF RESEARCH REACTOR SPENT FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D

    2006-08-22

    Aluminum clad research reactor spent nuclear fuel (SNF) is currently being consolidated in wet storage basins (pools). Approximately 20 metric tons (heavy metal) of aluminum-based spent nuclear fuel (Al-SNF) is being consolidated for treatment, packaging, interim storage, and preparation for ultimate disposal in a geologic repository. The storage and disposal of Al-SNF are subject to requirements that provide for safety and acceptable radionuclide release. The options studied for interim storage of SNF include wet storage and dry storage. Two options have also been studied to develop the technical basis for the qualification and repository disposal of aluminum spent fuel. The two options studied include Direct Disposal and Melt-Dilute treatment. The implementation of these options present relative benefits and challenges. Both the Direct Disposal and the Melt-Dilute treatment options have been developed and their technical viability assessed. Adaptation of the melt-dilute technology for the treatment of spent fuel offers the benefits of converting the spent fuel into a proliferation resistant form and/or significantly reducing the volume of the spent fuel. A Mobile Melt-Dilute system concept has emerged to realize these benefits and a prototype system developed. The application of the melt-dilute technology for the treatment of legacy nuclear materials has been evaluated and also offers the promise for the safe disposal of these materials.

  12. The Conceptual Design for a Fuel Assembly of a New Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, J-S.; Cho, Y-G.; Yoon, D-B.; Dan, H-J.; Chae, H-T.; Park, C.

    2004-10-06

    A new Research Reactor (ARR) has been under design by KAERI since 2002. In this work, as a first step for the design of the fuel assembly of the ARR, the conceptual design has been carried out. The vibration characteristics of the tubular fuel model and the locking performance of the preliminary designed locking devices were investigated. In order to investigate the effects of the stiffener on the vibration characteristics of the tubular fuel, a modal analysis was performed for the finite element models of the tubular fuels with stiffeners and without stiffeners. The analysis results show that the vibration characteristics of the tubular fuel with stiffeners are better than those of the tubular fuel without stiffeners. To investigate the locking performance of the preliminary designed locking devices for the fuel assembly of the ARR, the elements of the locking devices were fabricated. Then the torsional resistance, fixing status and vibration characteristics of the locking devices were tested. The test results show that using the locking device with fins on the bottom guide can prevent the torsional motion of the fuel assembly, and that additional springs or guides on the top of the fuel assembly are needed to suppress the lateral motion of the fuel assembly. Based on the modal analysis and experimental results, the fuel assembly and locking devices of the ARR were designed and its prototype was fabricated. The locking performance, pressure drop characteristics and vibration characteristics of the newly designed fuel assembly will be tested in the near future.

  13. Development of integrated management system for the research reactor in Sofia

    Energy Technology Data Exchange (ETDEWEB)

    Stoyanova, A.S.; Ilieva, K.T. [Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences, Sofia (Bulgaria)

    2007-07-01

    The main purpose to set up an Integrated Management System (IMS) is to guarantee the safety operation of nuclear facilities as well as to increase their exploitation effectiveness. To ensure the safety operation of nuclear facilities the Bulgarian Nuclear Regulatory Agency has created requirements and norms to prevent potential nuclear incidents, overdose irradiation or terrorist attacks opportunities. The IMS of the Institute for Nuclear Research and Nuclear Energy has been developed in a way to create an environment which guarantees the ways and means for: quality management according to ISO 9001:2000, environmental management according to ISO 14001:2004, management of safety requirements of the Bulgarian Nuclear Regulatory Agency, security and physical protection, and the management of the safe and health working conditions for the employees. The IMS is based on the concepts recommended by IAEA: the entirety of work can be structured and interpreted as a set of interacting processes that can be planned, performed, measured, assessed and improved, and, those performing assessing work, all contribute in achieving quality and ensuring safety. The IMS has been developed in the way to be continuously upgraded and additive. The IMS will be added with new instructions, procedures and others, which will correspond to new activities arising during the reactor reconstruction. The IMS was developed on the basis of state-of-the-art software ARIS in the way to achieve ease in communication, visualization, possibilities for assessment and continuous improvement of the IMS. (authors)

  14. Utilization of coal fly ash in solidification of liquid radioactive waste from research reactor.

    Science.gov (United States)

    Osmanlioglu, Ahmet Erdal

    2014-05-01

    In this study, the potential utilization of fly ash was investigated as an additive in solidification process of radioactive waste sludge from research reactor. Coal formations include various percentages of natural radioactive elements; therefore, coal fly ash includes various levels of radioactivity. For this reason, fly ashes have to be evaluated for potential environmental implications in case of further usage in any construction material. But for use in solidification of radioactive sludge, the radiological effects of fly ash are in the range of radioactive waste management limits. The results show that fly ash has a strong fixing capacity for radioactive isotopes. Specimens with addition of 5-15% fly ash to concrete was observed to be sufficient to achieve the target compressive strength of 20 MPa required for near-surface disposal. An optimum mixture comprising 15% fly ash, 35% cement, and 50% radioactive waste sludge could provide the solidification required for long-term storage and disposal. The codisposal of radioactive fly ash with radioactive sludge by solidification decreases the usage of cement in solidification process. By this method, radioactive fly ash can become a valuable additive instead of industrial waste. This study supports the utilization of fly ash in industry and the solidification of radioactive waste in the nuclear industry.

  15. Performance improvement of artificial neural networks designed for safety key parameters prediction in nuclear research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mazrou, Hakim [Division de Physique Radiologique, Centre de Recherche Nucleaire d' Alger (CRNA), 02 Boulevard Frantz, Fanon, B.P. 399, 16000 Alger (Algeria)], E-mail: mazrou_h@crna.dz

    2009-10-15

    The present work explores, through a comprehensive sensitivity study, a new methodology to find a suitable artificial neural network architecture which improves its performances capabilities in predicting two significant parameters in safety assessment i.e. the multiplication factor k{sub eff} and the fuel powers peaks P{sub max} of the benchmark 10 MW IAEA LEU core research reactor. The performances under consideration were the improvement of network predictions during the validation process and the speed up of computational time during the training phase. To reach this objective, we took benefit from Neural Network MATLAB Toolbox to carry out a widespread sensitivity study. Consequently, the speed up of several popular algorithms has been assessed during the training process. The comprehensive neural system was subsequently trained on different transfer functions, number of hidden neurons, levels of error and size of generalization corpus. Thus, using a personal computer with data created from preceding work, the final results obtained for the treated benchmark were improved in both network generalization phase and much more in computational time during the training process in comparison to the results obtained previously.

  16. Lewis Research Center's coal-fired, pressurized, fluidized-bed reactor test facility

    Science.gov (United States)

    Kobak, J. A.; Rollbuhler, R. J.

    1981-01-01

    A 200-kilowatt-thermal, pressurized, fluidized-bed (PFB) reactor, research test facility was designed, constructed, and operated as part of a NASA-funded project to assess and evaluate the effect of PFB hot-gas effluent on aircraft turbine engine materials that might have applications in stationary-power-plant turbogenerators. Some of the techniques and components developed for this PFB system are described. One of the more important items was the development of a two-in-one, gas-solids separator that removed 95+ percent of the solids in 1600 F to 1900 F gases. Another was a coal and sorbent feed and mixing system for injecting the fuel into the pressurized combustor. Also important were the controls and data-acquisition systems that enabled one person to operate the entire facility. The solid, liquid, and gas sub-systems all had problems that were solved over the 2-year operating time of the facility, which culminated in a 400-hour, hot-gas, turbine test.

  17. Fast neutron radiography and tomography at a 10MW research reactor beamline.

    Science.gov (United States)

    Zboray, R; Adams, R; Kis, Z

    2017-01-01

    Fast neutron imaging was performed using a beamline of the 10MW research reactor of the Budapest Neutron Centre, Hungary. A simple, low-cost 2D area detector has been used featuring a 8mm thick BC400 plastic scintillator converter screen and a CCD camera. A spatial resolution of around 1.3mm has been achieved. Typically 10min long exposures were needed to obtain reasonable quality radiographic images. For tomographic imaging typically several hours of acquisition were needed to obtain reasonable quality on non-symmetric and larger (e.g. 10×10×10cm3) objects. Due to the presence of a significant gamma background at the experimental position, massive (30cm thick) lead shielding and filtering was applied to the beam. The gamma contribution was mostly baseline independent of the object imaged and therefore could be subtracted, whereas the direct gamma contribution from the beam to the imaging detector signal is estimated to be less than 1%. Copyright © 2016 Elsevier Ltd. All rights reserved.

  18. A new small-angle neutron scattering spectrometer at China Mianyang research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Mei, E-mail: pm740509@163.com; Sun, Liangwei; Chen, Liang; Sun, Guangai; Chen, Bo; Xie, Chaomei; Xia, Qingzhong; Yan, Guanyun; Tian, Qiang; Huang, Chaoqiang; Pang, Beibei; Zhang, Ying; Wang, Yun; Liu, Yaoguang; Kang, Wu; Gong, Jian

    2016-02-21

    A new pinhole small-angle neutron scattering (SANS) spectrometer, installed at the cold neutron source of the 20 MW China Mianyang Research Reactor (CMRR) in the Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, has been put into use since 2014. The spectrometer is equipped with a multi-blade mechanical velocity selector, a multi-beam collimation system, and a two-dimensional He-3 position sensitive neutron detector. The q-range of the spectrometer covers from 0.01 nm{sup −1} to 5.0 nm{sup −1}. In this paper, the design and characteristics of the SANS spectrometer are described. The q-resolution calculations, together with calibration measurements of silver behenate and a dispersion of nearly monodisperse poly-methyl-methacrylate nanoparticles indicate that our SANS spectrometer has a good performance and is now in routine service. - Highlights: • A new SANS spectrometer has been put into use since 2014 in China. • One MBR selector possesses a higher resolution compared with traditional selector is used. • The spectrometer has a good performance and is now in routinely service.

  19. Study on disposal method of graphite blocks and storage of spent fuel for modular gas-cooled reactor. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Sumita, Junya; Sawa, Kazuhiro; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tsuchie, Yasuo; Urakami, Masao [Japan Atomic Power Co., Tokyo (Japan)

    2003-02-01

    This report describes the result of study on disposal method of graphite blocks in future block-type reactor. Present study was carried out within a framework of joint research, ''Research of Modular High Temperature Gas-cooled Reactors (No. 3)'', between Japan Atomic Energy Research Institute (JAERI) and the Japan Atomic Power Company (JAPCO), in 2000. In this study, activities in fuel and reflector graphite blocks were evaluated and were compared with the disposal limits defined as low-level of radioactive waste. As a result, it was found that the activity for only C-14 was higher than disposal limits for the low-level of radioactive waste and that the amount of air in the graphite is important to evaluate precisely of C-14 activity. In addition, spent fuels can be stored in air-cooled condition at least after two years cooling in the storage pool. (author)

  20. Advanced reactor safety research quarterly report, October-December 1982. Volume 24

    Energy Technology Data Exchange (ETDEWEB)

    None

    1984-04-01

    This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  1. [Research on change process of nitrosation granular sludge in continuous stirred-tank reactor].

    Science.gov (United States)

    Yin, Fang-Fang; Liu, Wen-Ru; Wang, Jian-Fang; Wu, Peng; Shen, Yao-Liang

    2014-11-01

    In order to investigate the effect of different types of reactors on the nitrosation granular sludge, a continuous stirred-tank reactor (CSTR) was studied, using mature nitrosation granular sludge cultivated in sequencing batch reactor (SBR) as seed sludge. Results indicated that the change of reactor type and influent mode could induce part of granules to lose stability with gradual decrease in sludge settling ability during the initial period of operation. However, the flocs in CSTR achieved fast granulation in the following reactor operation. In spite of the changes of particle size distribution, e. g. the decreasing number of granules with diameter larger than 2.5 mm and the increasing number of granules with diameter smaller than 0.3 mm, granular sludge held the absolute predominance of sludge morphology in CSTR during the entire experimental period. Moreover, results showed that the change of reactor type and influent mode didn't affect the nitrite accumulation rate which was still kept at about 85% in effluent. Additionally, the average activity of the sludge in CSTR was stronger than that of the seed sludge, because the newly generated small particles in CSTR had higher specific reactive activity than the larger granules.

  2. Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

    1994-10-01

    This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

  3. Light Water Reactor Sustainability Research and Development Program Plan -- Fiscal Year 2009–2013

    Energy Technology Data Exchange (ETDEWEB)

    Idaho National Laboratory

    2009-12-01

    Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R&D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R&D programs. The purpose of the LWRS R&D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R&D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its international engagement

  4. Feynman-alpha technique for measurement of detector dead time using a 30 kW tank-in-pool research reactor

    CERN Document Server

    Akaho, E H K; Intsiful, J D K; Maakuu, B T; Nyarko, B J B

    2002-01-01

    Reactor noise analysis was carried out for Ghana Research Reactor-1 GHARR-1, a tank-in-pool type reactor using the Feynman-alpha technique (variance-to-mean method). Measurements made at different detector positions and under subcritical conditions showed that the technique could not be used to determine the prompt decay constant for the reactor which is Be reflected with photo-neutron background. However, for very low dwell times the technique was used to measure the dead time of the detector which compares favourably with the value obtained using the alpha-conventional method.

  5. Soil bed reactor work of the Environmental Research Lab. of the University of Arizona in support of the research and development of Biosphere 2

    Science.gov (United States)

    Frye, Robert

    1990-01-01

    Research at the Environmental Research Lab in support of Biosphere 2 was both basic and applied in nature. One aspect of the applied research involved the use of biological reactors for the scrubbing of trace atmospheric organic contaminants. The research involved a quantitative study of the efficiency of operation of Soil Bed Reactors (SBR) and the optimal operating conditions for contaminant removal. The basic configuration of a SBR is that air is moved through a living soil that supports a population of plants. Upon exposure to the soil, contaminants are either passively adsorbed onto the surface of soil particles, chemically transformed in the soil to usable compounds that are taken up by the plants or microbes as a metabolic energy source and converted to CO2 and water.

  6. Research and development on the application of advanced control technologies to advanced nuclear reactor systems: A US national perspective

    Energy Technology Data Exchange (ETDEWEB)

    White, J.D.; Monson, L.R.; Carrol, D.G.; Dayal, Y. (Oak Ridge National Lab., TN (USA); Argonne National Lab., IL (USA); General Electric Co., San Jose, CA (USA))

    1989-01-01

    Control system designs for nuclear power plants are becoming more advanced through the use of digital technology and automation. This evolution is taking place because of: (1) the limitations in analog based control system performance and maintenance and availability and (2) the promise of significant improvement in plant operation and availability due to advances in digital and other control technologies. Digital retrofits of control systems in US nuclear plants are occurring now. Designs of control and protection systems for advanced LWRs are based on digital technology. The use of small inexpensive, fast, large-capacity computers in these designs is the first step of an evolutionary process described in this paper. Under the sponsorship of the US Department of Energy (DOE), Oak Ridge National Laboratory, Argonne National Laboratory, GE Nuclear Energy and several universities are performing research and development in the application of advances in control theory, software engineering, advanced computer architectures, artificial intelligence, and man-machine interface analysis to control system design. The target plant concept for the work described in this paper is the Power Reactor Inherently Safe Module reactor (PRISM), an advanced modular liquid metal reactor concept. This and other reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. 18 refs., 5 figs.

  7. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY1999-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    The HTTR (High Temperature Engineering Test Reactor) with the thermal power of 30 MW and the reactor outlet coolant temperature of 850/950 degC is the first high temperature gas-cooled reactor (HTGR) in Japan, which uses coated fuel particle, graphite for core components, and helium gas for primary coolant. The HTTR, which locates at the south-west area of 50,000 m{sup 2} in the Oarai Research Establishment, had been constructed since 1991 before accomplishing the first criticality on November 10, 1998. Rise to power tests of the HTTR started in September, 1999 and the rated thermal power of 30 MW and the reactor outlet coolant temperature of 850 degC was attained in December 2001. JAERI received the certificate of pre-operation test, that is, the commissioning license for the HTTR in March 2002. This report summarizes operation, tests, maintenance, radiation control, and construction of components and facilities for the HTTR as well as R and Ds on HTGRs from FY1999 to 2001. (author)

  8. Experimental evaluation of gamma fluence-rate predictions from Argon-41 releases to the atmosphere over a nuclear research reactor site

    DEFF Research Database (Denmark)

    Rojas-Palma, C.; Aage, H.K.; Astrup, P.

    2004-01-01

    An experimental study of radionuclide dispersion in the atmosphere has been conducted at the BR1 research reactor in Mol, Belgium. Artificially generated aerosols ('white smoke') were mixed with the routine releases of Ar-41 in the reactor's 60-m tall venting stack. The detailed plume geometry...

  9. Impact of mesh points number on the accuracy of deterministic calculations of control rods worth for Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Boustani, Ehsan [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Amirkabir University of Technology, Tehran (Iran, Islamic Republic of). Energy Engineering and Physics Dept.; Khakshournia, Samad [Amirkabir University of Technology, Tehran (Iran, Islamic Republic of). Energy Engineering and Physics Dept.

    2016-12-15

    In this paper two different computational approaches, a deterministic and a stochastic one, were used for calculation of the control rods worth of the Tehran research reactor. For the deterministic approach the MTRPC package composed of the WIMS code and diffusion code CITVAP was used, while for the stochastic one the Monte Carlo code MCNPX was applied. On comparing our results obtained by the Monte Carlo approach and those previously reported in the Safety Analysis Report (SAR) of Tehran research reactor produced by the deterministic approach large discrepancies were seen. To uncover the root cause of these discrepancies, some efforts were made and finally was discerned that the number of spatial mesh points in the deterministic approach was the critical cause of these discrepancies. Therefore, the mesh optimization was performed for different regions of the core such that the results of deterministic approach based on the optimized mesh points have a good agreement with those obtained by the Monte Carlo approach.

  10. Atmospheric dispersion of argon-41 from anuclear research reactor: measurement and modeling of plume geometry and gamma radiation field

    DEFF Research Database (Denmark)

    Lauritzen, Bent; Astrup, Poul; Drews, Martin

    2003-01-01

    An atmospheric dispersion experiment was conducted using a visible tracer along with the routine release of argon-41 from the BR1 research reactor in Mol, Belgium. Simultaneous measurements of plume geometry and radiation fields for argon-41 decay were performed as well as measurements of the argon......-41 source term and the meteorological parametres. Good overall agreement is found between measurement data and model results using the mesoscale atmospheric dispersion and dose rate model RIMPUFF....

  11. Releasable activity and maximum permissible leakage rate within a transport cask of Tehran Research Reactor fuel samples

    OpenAIRE

    Rezaeian Mahdi; Kamali Jamshid; Roshanzamir Manoochehr; Moosakhani Alireza; Noori Elghar

    2015-01-01

    Containment of a transport cask during both normal and accident conditions is important to the health and safety of the public and of the operators. Based on IAEA regulations, releasable activity and maximum permissible volumetric leakage rate within the cask containing fuel samples of Tehran Research Reactor enclosed in an irradiated capsule are calculated. The contributions to the total activity from the four sources of gas, volatile, fines, and corrosion...

  12. Feasibility studies of producing {sup 99} Mo by capture in the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Concilio, Roberta; Mendonca, Arlindo Gilson; Maiorino, Jose Rubens [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: maiorino@net.ipen.br; amendon@net.ipen.br

    1998-07-01

    Everyday the production of {sup 99} Mo for {sup 99m} Tc generators, becomes more necessary, whose properties are ideal for medical diagnosis. This works presents a description and an analysis of the production of {sup 99} Mo by radioactive capture at {sup 98} Mo using the research reactor IEA-R1 in 5 MW and operating 5 days a week, referring to the use of targets, separation methods, total and specific activity attained and its limitations. (author)

  13. Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

    2012-09-30

    Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.

  14. TAIPAN: First Results from the Thermal Triple-axis Spectrometer at OPAL Research Reactor

    Science.gov (United States)

    Danilkin, S. A.; Yethiraj, M.; Saerbeck, T.; Klose, F.; Ulrich, C.; Fujioka, J.; Miyasaka, S.; Tokura, Y.; Keimer, B.

    2012-02-01

    The thermal triple-axis spectrometer TAIPAN is the first instrument for inelastic neutron scattering at the new Australian research reactor OPAL. TAIPAN started operation in February 2009 and is in full user service since November 2010. Conceptually, it is similar to the triple-axis spectrometers IN8 (ILL) and PANDA (FRM-II) with variable incident and final energies and a secondary spectrometer with a single detector. The instrument can be operated either in a high flux mode with a double-focusing monochromator and analyser, or with Soller collimators - gaining resolution at the expense of intensity. Presently the PG (002) double-focusing monochromator and analyser are in use. The incident energy range on the TAIPAN TAS is from ~5 meV up to ~100 meV with neutron flux at sample position of 2.4<=107 n/cm2/s at incident energy of 14.8 meV. First experiments were performed with superionic conductor Cu2-δSe. The measurements reveal the presence of a soft mode related to ordering of Cu atoms followed by α - β phase transition at a lower temperature. The evolution of the magnetic structure with temperature in a magnetically modulated FePt3 thin film was investigated in the diffraction mode of TAIPAN. The results show that the film fabricated by modulation of the chemical order parameter consists of a magnetic FM/AFM superlattice in single-crystalline FePt3. The spin wave and phonon dispersion was recently investigated in TbVO3 single crystal. The acoustic and optical magnon branches were observed in the same energy range. This indicates that the 'orbital Peiers state' also exists in TbVO3.

  15. Pressure Loss Predictions of the Reactor Simulator Subsystem at NASA Glenn Research Center

    Science.gov (United States)

    Reid, Terry V.

    2016-01-01

    Testing of the Fission Power System (FPS) Technology Demonstration Unit (TDU) is being conducted at NASA Glenn Research Center. The TDU consists of three subsystems: the reactor simulator (RxSim), the Stirling Power Conversion Unit (PCU), and the heat exchanger manifold (HXM). An annular linear induction pump (ALIP) is used to drive the working fluid. A preliminary version of the TDU system (which excludes the PCU for now) is referred to as the "RxSim subsystem" and was used to conduct flow tests in Vacuum Facility 6 (VF 6). In parallel, a computational model of the RxSim subsystem was created based on the computer-aided-design (CAD) model and was used to predict loop pressure losses over a range of mass flows. This was done to assess the ability of the pump to meet the design intent mass flow demand. Measured data indicates that the pump can produce 2.333 kg/sec of flow, which is enough to supply the RxSim subsystem with a nominal flow of 1.75 kg/sec. Computational predictions indicated that the pump could provide 2.157 kg/sec (using the Spalart-Allmaras (S?A) turbulence model) and 2.223 kg/sec (using the k- turbulence model). The computational error of the predictions for the available mass flow is ?0.176 kg/sec (with the S-A turbulence model) and -0.110 kg/sec (with the k- turbulence model) when compared to measured data.

  16. Development of a New Transportation/Storage Cask System for Use by the DOE Russian Research Reactor Fuel Return Program

    Energy Technology Data Exchange (ETDEWEB)

    Michael J. Tyacke; Frantisek Svitak; Jiri Rychecky; Miroslav Picek; Alexey Smirnov; Sergey Komarov; Edward Bradley; Alexander Dudchenko; Konstantin Golubkin

    2007-10-01

    The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions at these research reactor facilities. This new cask system is named the ŠKODA VPVR/M cask. The design, licensing, testing, and delivery of this new cask system result from a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: 1) Introduction; 2) VPVR/M Cask Description; 3) Ancillary Equipment, 4) Cask Licensing; 5) Cask Demonstration and Operations; 6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, 7) Conclusions.

  17. Fuel irradiation research of Japan at OECD Halden Reactor Project. Achievement of joint researches between JAERI and other organizations in the period from 1994 to 1996

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi; Nakamura, Jinichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kinoshita, Motoyasu [and others

    1998-01-01

    JAERI has performed cooperative researches with many Japanese agencies and companies by means of the Halden Boiling Heavy Water Reactor (HBWR) which is located at Halden in Norway. These cooperative researches are carried out based on the contracts of the cooperative researches, which are revised every three years, in accordance with the renewal of the participation of JAERI to the OECD Halden Reactor Project. This report summaries the objectives, contents and the outlines of the achievements of the cooperative researches during the three years from 1994 January to 1996 December. During the period, ten cooperative researches had been carried out, and two of them had finished during the period and other eight researches has been continued to the next three year period. There are many research items, and most of them are irradiation test researches of advanced fuel and cladding concerned with the high burnup utilization of LWR fuel or MOX fuel irradiation researches to prepare for the introduction of Plutonium utilization in LWRs. The researches of fuel irradiation usually take long time because of the characteristics of these kind of research work, and three years are usually not enough to obtain some achievements from the irradiation tests. Therefore, eight tests have been continued after the three year period. In this report, the achievements of the continued researches to the next three year period are not final one but a kind of progress report. (author)

  18. Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T [ORNL

    2014-05-01

    Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

  19. Evaluation of integral continuing experimental capability (CEC) concepts for light water reactor research: PWR scaling concepts

    Energy Technology Data Exchange (ETDEWEB)

    Condie, K G; Larson, T K; Davis, C B; McCreery, G E

    1987-02-01

    In this report reactor transients and thermal-hydraulic phenomena of importance (based on probabilistic risk assessment and the International Code Assessment Program) to reactor safety were examined and identified. Established scaling methodologies were used to develop potential concepts for integral thermal-hydraulic testing facilities. Advantages and disadvantages of each concept are evaluated. Analysis is conducted to examine the scaling of various phenomena in each of the selected concepts. Results generally suggest that a facility capable of operating at typical reactor operating conditions will scale most phenomena reasonably well. Although many phenomena in facilities using Freon or water at nontypical pressure will scale reasonably well, those phenomena that are heavily dependent on quality (heat transfer or critical flow for example) can be distorted. Furthermore, relation of data produced in facilities operating with nontypical fluids or at nontypical pressures to large plants will be a difficult and time consuming process.

  20. International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

    2014-12-15

    Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

  1. Research on soybean protein wastewater treatment by the integrated two-phase anaerobic reactor

    Science.gov (United States)

    Yu, Yaqin

    2015-01-01

    The start-up tests of treating soybean protein wastewater by the integrated two-phase anaerobic reactor were studied. The results showed that the soybean protein wastewater could be successfully processed around 30 days when running under the situation of dosing seed sludge with the influent of approximately 2000 mg/L and an HRT of 40 h. When the start-up was finished, the removal rate of COD by the reactor was about 80%. In the zone I, biogas mainly revealed carbon dioxide (CO2) and hydrogen (H2). Methane was the main component in the zone 2 which ranged from 53% to 59% with an average of 55%. The methane content in biogas increased from the zone I to II. It indicated that the methane-producing capacity of the anaerobic sludge increased. It was found that the uniquely designed two-phase integrated anaerobic reactor played a key role in treating soybean protein wastewater. The acidogenic fermentation bacteria dominated in the zone I, while methanogen became dominant in the zone II. It realized the relatively effective separation of hydrolysis acidification and methanogenesis process in the reactor, which was benefit to promote a more reasonable space distribution of the microbial communities in the reactor. There were some differences between the activities of the sludge in the two reaction zones of the integrated two-phase anaerobic reactor. The activity of protease was higher in the reaction zone I. And the coenzyme F420 in the reaction zone II was twice than that in the reaction zone I, which indicated that the activity of the methanogens was stronger in the reaction zone II. PMID:26288554

  2. Research on soybean protein wastewater treatment by the integrated two-phase anaerobic reactor.

    Science.gov (United States)

    Yu, Yaqin

    2015-09-01

    The start-up tests of treating soybean protein wastewater by the integrated two-phase anaerobic reactor were studied. The results showed that the soybean protein wastewater could be successfully processed around 30 days when running under the situation of dosing seed sludge with the influent of approximately 2000 mg/L and an HRT of 40 h. When the start-up was finished, the removal rate of COD by the reactor was about 80%. In the zone I, biogas mainly revealed carbon dioxide (CO2) and hydrogen (H2). Methane was the main component in the zone 2 which ranged from 53% to 59% with an average of 55%. The methane content in biogas increased from the zone I to II. It indicated that the methane-producing capacity of the anaerobic sludge increased. It was found that the uniquely designed two-phase integrated anaerobic reactor played a key role in treating soybean protein wastewater. The acidogenic fermentation bacteria dominated in the zone I, while methanogen became dominant in the zone II. It realized the relatively effective separation of hydrolysis acidification and methanogenesis process in the reactor, which was benefit to promote a more reasonable space distribution of the microbial communities in the reactor. There were some differences between the activities of the sludge in the two reaction zones of the integrated two-phase anaerobic reactor. The activity of protease was higher in the reaction zone I. And the coenzyme F420 in the reaction zone II was twice than that in the reaction zone I, which indicated that the activity of the methanogens was stronger in the reaction zone II.

  3. In-Vessel Melt Retention of Pressurized Water Reactors: Historical Review and Future Research Needs

    OpenAIRE

    Ma, Weimin; Yuan, Yidan; Sehgal, Bal Raj

    2016-01-01

    A historical review of in-vessel melt retention (IVR) is given, which is a severe accident mitigation measure extensively applied in Generation III pressurized water reactors (PWRs). The idea of IVR actually originated from the back-fitting of the Generation II reactor Loviisa VVER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse AP1000, the Korean APR1400 as well as Chinese advanced PWR designs HPR1000 and CAP1400. The most influential...

  4. Ageing Management of Beryllium and Graphite Blocks in Research Reactor MARIA

    Energy Technology Data Exchange (ETDEWEB)

    Golab, A. [National Centre for Nuclear Research, Warsaw (Poland)

    2013-07-01

    In the paper the phenomenon of beryllium moderator poisoning by thermal neutron absorption and the method and results of this phenomenon control is presented. Also the phenomenon of graphite blocks damage due to fast neutrons accumulation and the methods and results of this process supervising is described. These methods refer especially to: visual inspection of their state and radiography of graphite blocks. Special attention is paid to permanent estimate of fast neutron fluency accumulated in blocks and methods of their shuffling in the reactor core. The shuffling makes possible to increase the lifetime of beryllium and graphite blocks and decrease the cost of reactor operation.

  5. A comparative assessment of independent thermal-hydraulic models for research reactors: The RSG-GAS case

    Energy Technology Data Exchange (ETDEWEB)

    Chatzidakis, S., E-mail: schatzid@purdue.edu [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907 (United States); Hainoun, A. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400, San Carlos de Bariloche, Rio Negro (Argentina); Alhabet, F. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Francioni, F. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400, San Carlos de Bariloche, Rio Negro (Argentina); Ikonomopoulos, A. [Institute of Nuclear and Radiological Sciences, Energy, Technology and Safety, National Center for Scientific Research ‘Demokritos’, 15130, Aghia Paraskevi, Athens (Greece); Ridikas, D. [Department of Nuclear Sciences and Applications, International Atomic Energy Agency, Vienna International Centre, A-1400 Vienna (Austria)

    2014-03-15

    Highlights: • Increased use of thermal-hydraulic codes requires assessment of important phenomena in RRs. • Three independent modeling teams performed analysis of loss of flow transient. • Purpose of this work is to examine the thermal-hydraulic codes response. • To perform benchmark analysis comparing the different codes with experimental measurements. • To identify the impact of the user effect on the computed results, performed with the same codes. - Abstract: This study presents the comparative assessment of three thermal-hydraulic codes employed to model the Indonesian research reactor (RSG-GAS) and simulate the reactor behavior under steady state and loss of flow transient (LOFT). The RELAP5/MOD3, MERSAT and PARET-ANL thermal-hydraulic codes are used by independent research groups to perform benchmark analysis against measurements of coolant and clad temperatures, conducted on an instrumented fuel element inside RSG-GAS core. The results obtained confirm the applicability of RELAP5/MOD3, MERSAT and PARET-ANL on the modeling of loss of flow transient in research reactors. In particular, the three codes are able to simulate flow reversal from downward forced to upward natural convection after pump trip and successful reactor scram. The benchmark results show that the codes predict maximum clad temperature of hot channel conservatively with a maximum overestimation of 27% for RELAP5/MOD3, 17% for MERSAT and 8% for PARET-ANL. As an additional effort, the impact of user effect on the simulation results has been assessed for the code RELAP5/MOD3, where the main differences among the models are presented and discussed.

  6. History of the research reactor institute of Kyoto University in view of nuclear science information data base (KURRIP)

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, Takayuki; Mizuma, Mitsuo (Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.); Kimura, Itsuro

    1994-02-01

    Since the Research Reactor Institute of Kyoto University was established as an inter-university research institute in 1963, a large number of cooperative research projects have been achieved by visiting scientists and its own staff in various research fields, making use of facilities centered around the Kyoto University Reactor, as well as the other experimental facilities. Ten years ago, the construction of the 'KURRIP' data base was initiated to grasp the whole aspect of the research activities at the Institute, in commemoration of its 20th anniversary. At the present time, KURRIP contains the information on 5,910 papers published for 29 years from 1963 to 1991. As this academic year is the 30th anniversary of the Institute, the history of its research activities was reviewed again using this data base. All of the publications were classified by authors's affiliations, kinds of papers, publishers, fields of studies, and research facilities used, and their historical variations are checked and discussed. (author).

  7. Development of Pneumatic Transfer Irradiation Facility (PTS no.3) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-04-15

    A pneumatic transfer system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide. The pneumatic transfer irradiation system (PTS no.3) involving a manual system and an semi-automatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor and NAA laboratory of RI building in 2006. In this technical report, the design, operation and control of these system (PTS no.3) was described. Also the experimental results and the characteristic parameters measured from a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  8. Development of Pneumatic Transfer Irradiation Facility (PTS no.1) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-03-15

    A pneumatic transfer system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide and a delayed neutron counting system. The pneumatic transfer system (PTS no.1) involving a manual system and an semiautomatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. In this technical report, the conception, design, operation and control of these system (PTS no.1) was described. Also the experimental results and the characteristic parameters measured by a mock-up test, a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  9. Development of Pneumatic Transfer Irradiation Facility (PTS no.2) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-03-15

    A pneumatic transfer irradiation system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide and a delayed neutron counting system. The pneumatic transfer irradiation system (PTS no.2) involving a manual system and an automatic system for delayed neutron activation analysis (DNAA) were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. In this technical report, the conception, design, operation and control of PTS no.2 was described. Also the experimental results and the characteristic parameters measured by a mock-up test, a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, automatic operation control by personal computer, delayed neutron counting system, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  10. Prediction of Flow and Temperature Distributions in a High Flux Research Reactor Using the Porous Media Approach

    Directory of Open Access Journals (Sweden)

    Shanfang Huang

    2017-01-01

    Full Text Available High thermal neutron fluxes are needed in some research reactors and for irradiation tests of materials. A High Flux Research Reactor (HFRR with an inverse flux trap-converter target structure is being developed by the Reactor Engineering Analysis Lab (REAL at Tsinghua University. This paper studies the safety of the HFRR core by full core flow and temperature calculations using the porous media approach. The thermal nonequilibrium model is used in the porous media energy equation to calculate coolant and fuel assembly temperatures separately. The calculation results show that the coolant temperature keeps increasing along the flow direction, while the fuel temperature increases first and decreases afterwards. As long as the inlet coolant mass flow rate is greater than 450 kg/s, the peak cladding temperatures in the fuel assemblies are lower than the local saturation temperatures and no boiling exists. The flow distribution in the core is homogeneous with a small flow rate variation less than 5% for different assemblies. A large recirculation zone is observed in the outlet region. Moreover, the porous media model is compared with the exact model and found to be much more efficient than a detailed simulation of all the core components.

  11. Derivation of the source term, dose results and associated radiological consequences for the Greek Research Reactor – 1

    Energy Technology Data Exchange (ETDEWEB)

    Pappas, Charalampos, E-mail: chpappas@ipta.demokritos.gr; Ikonomopoulos, Andreas; Sfetsos, Athanasios; Andronopoulos, Spyros; Varvayanni, Melpomeni; Catsaros, Nicolas

    2014-07-01

    Highlights: • Source term derivation of postulated accident sequences in a research reactor. • Various containment ventilation scenarios considered for source term calculations. • Source term parametric analysis performed in case of lack of ventilation. • JRODOS employed for dose calculations under eighteen modeled scenarios. • Estimation of radiological consequences during typical and adverse weather scenarios. - Abstract: The estimated source term, dose results and radiological consequences of selected accident sequences in the Greek Research Reactor – 1 are presented and discussed. A systematic approach has been adopted to perform the necessary calculations in accordance with the latest computational developments and IAEA recommendations. Loss-of-coolant, reactivity insertion and fuel channel blockage accident sequences have been selected to derive the associated source terms under three distinct containment ventilation scenarios. Core damage has been conservatively assessed for each accident sequence while the ventilation has been assumed to function within the efficiency limits defined at the Safety Analysis Report. In case of lack of ventilation a parametric analysis is also performed to examine the dependency of the source term on the containment leakage rate. A typical as well as an adverse meteorological scenario have been defined in the JRODOS computational platform in order to predict the effective, lung and thyroid doses within a region defined by a 15 km radius downwind from the reactor building. The radiological consequences of the eighteen scenarios associated with the accident sequences are presented and discussed.

  12. Safety evaluation report related to the renewal of the operating license for the research reactor at North Carolina State University

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

    This safety evaluation report (SER) summarizes the findings of a safety review conducted by the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation (NRR). The staff conducted this review in response to a timely application filed by North Carolina State University (the licensee or NCSU) for a 20-year renewal of Facility Operating License R-120 to continue to operate the NCSU PULSTAR research reactor. The facility is located in the Burlington Engineering Laboratory complex on the NCSU campus in Raleigh, North Carolina. In its safety review, the staff considered information submitted by the licensee (including past operating history recorded in the licensee`s annual reports to the NRC), as well as inspection reports prepared by NRC Region H personnel and first-hand observations. On the basis of this review, the staff concludes that NCSU can continue to operate the PULSTAR research reactor, in accordance with its application, without endangering the health and safety of the public. 16 refs., 31 figs., 7 tabs.

  13. Membrane bio-reactor - Research, pilot installation and measurement campaign; Membranbioreaktor (MBR) - Forschung, Pilotanlage und Messkampagne - Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Hersener, J.-L. [Ingenieurbuero Hersener, Wiesendangen (Switzerland); Meier, U. [Meritec GmbH, Guntershausen (Switzerland)

    2007-07-01

    This report for the Swiss Federal Office of Energy (SFOE), takes a look at a project involving a fermenter installation in Eastern Switzerland. Research work is noted, the pilot installation is described and the results of a measurement campaign are presented and commented on. The plant is able to handle about 20,000-25,000 tonnes of slurry and organic waste. The plant is built as a membrane bio-reactor and allows the separation of the digested biomass into fractions of solid and liquid fertilisers and useful water. Furthermore, a part of the separated and digested liquid is returned to the fermenter in order to improve the digestion process. For the production of electricity a 1.1 MW generator is installed. The adaptations made during the measurement period are noted and commented on. According to the authors, the results - although difficult to interpret - show that the concept of a membrane bio-reactor can work successfully.

  14. Radiation and criticality safety analyses for the highly-enriched uranium core removal from a research reactor.

    Science.gov (United States)

    Dennis, Haile; Grant, Charles; Preston, John

    2017-11-01

    Analysis was performed to estimate radiation levels during removal and packaging of the highly-enriched uranium core of the JM-1 SLOWPOKE-2 research reactor. Due to severe limitations of space in and around the reactor pool, the core could not be removed in the conventional manner as was done for previous SLOWPOKE defuelling operations. A transfer shield, with a balance between shielding efficacy, volume and weight was designed. Fuel depletion, Monte Carlo shielding and criticality calculations were performed. Comparisons of measured and calculated dose rates as well as results of the criticality safety assessment are presented. The designed transfer shield reduced the calculated unshielded dose rate from 29Sv/h to 8mSv/h. The maximum calculated effective neutron multiplication factor of approximately 0.89 was below the 0.91 upper subricital limit. Copyright © 2017 Elsevier Ltd. All rights reserved.

  15. Gaseous-fuel nuclear reactor research for multimegawatt power in space

    Science.gov (United States)

    Thom, K.; Schneider, R. T.; Helmick, H. H.

    1977-01-01

    In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

  16. Advanced reactor safety research. Quarterly report, April-June 1982. Volume 22

    Energy Technology Data Exchange (ETDEWEB)

    None

    1983-10-01

    Overall objective of this work is to provide NRC a comprehensive data base essential to (1) defining key safety issues, (2) understanding risk-significant accident sequences, (3) developing and verifying models used in safety assessments, and (4) assuring the public that power reactor systems will not be licensed and placed in commercial service in the United States without appropriate consideration being given to their effects on health and safety. This report describes progress in a number of activities dealing with current safety issues relevant to both light water and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents, and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  17. Neutronic simulation of a research reactor core of (232 Th, 235 U ...

    Indian Academy of Sciences (India)

    A simple nuclear reactor core arranged 19 assemblies in hexagonal structure, using thorium-based fuel and heavy water as coolant and moderator was simulated using MCNPX2.6 code, aiming an optimized critical assembly. Optimized reflector thickness and gap between assemblies were determined to achieve minimum ...

  18. Neutron spectrum measurements at a radial beam port of the NUR research reactor using a Bonner spheres spectrometer.

    Science.gov (United States)

    Mazrou, H; Nedjar, A; Seguini, T

    2016-08-01

    This paper describes the measurement campaign held around the neutron radiography (NR) facility of the Algerian 1MW NUR research reactor. The main objective of this work is to characterize accurately the neutron beam provided at one of the radial channels of the NUR research reactor taking benefit of the acquired CRNA Bonner spheres spectrometer (BSS). The specific objective was to improve the image quality of the NR facility. The spectrometric system in use is based on a central spherical (3)He thermal neutron proportional counter combined with high density polyethylene spheres of different diameters ranging from 3 to 12in. This counting system has good gamma ray discrimination and is able to cover an energy range from thermal to 20MeV. The measurements were performed at the sample distance of 0.6m from the beam port and at a height of 1.2m from the facility floor. During the BSS measurements, the reactor was operating at low power (100W) to avoid large dead times, pulse pileup and high level radiation exposures, in particular, during spheres handling. Thereafter, the neutron spectrum at the sample position was unfolded by means of GRAVEL and MAXED computer codes. The thermal, epithermal and fast neutron fluxes, the total neutron flux, the mean energy and the Cadmium ratio (RCd) were provided. A sensitivity analysis was performed taking into account various defaults spectra and ultimately a different response functions in the unfolding procedure. Overall, from the obtained results it reveals, unexpectedly, that the measured neutron spectrum at the sample position of the neutron radiography of the NUR reactor is being harder with a predominance of fast neutrons (>100keV) by about 60%. Finally, those results were compared to previous and more recent measurements obtained by activation foils detectors. The agreement was fairly good highlighting thereby the consistency of our findings. Copyright © 2016 Elsevier Ltd. All rights reserved.

  19. Report of the second joint Research Committee for Fusion Reactor and Materials. July 12, 2002, Tokyo, Japan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    Joint research committees in purpose of the discussion on DEMO blanket in view point of the both of reactor technology and materials were held by the Research Committee for Fusion Reactor and Fusion Materials. The joint research committee was held in Tokyo on July 12, 2002. In the committee, the present status of development of solid and liquid breeding blanket, the present status of development of reduced activation structure materials, and IFMIF (International Fusion Materials Irradiation Facility) program were discussed based on the discussions of the development programs of the blanket and materials at the first joint research committee. As a result, it was confirmed that high electric efficiency with 41% would be obtained in the solid breeding blanket system, that neutron radiation data of reduced activation ferritic steel was obtained by HFIR collaboration, and that KEP (key element technology phase) of IFMIF would be finished at the end of 2002 and the data base for the next step, i.e. EVEDA (engineering validation/engineering design activity) was obtained. In addition, the present status of ITER CTA, which was a transient phase for the construction, and the outline of ITER Fast Track, which was an accelerated plan for the performance of the power plants, were reported. This report consists of the summary of the discussion and the viewgraphs which were used at the second joint research committee, and these are very useful for the researchers of the fusion area in Japan. (author)

  20. Testing and Research Capabilities at the Sandia Fast Pulsed Reactor Facility

    Science.gov (United States)

    Berry, Donald T.

    1994-07-01

    A wide variety of space-based system components have been qualified for use through neutron irradiation testing performed at the Sandia Pulsed Reactor (SPR) Facility. The SPR Facility is the operating location for two fast burst reactors, SPR II and SPR III, which have been used to induce neutron and gamma damage in electronic components and other materials for customers in the Department of Energy, Department of Defense, NASA, and the private sector. In addition to the pulse mode of operation, during which peak fluxes of up to 1023 n/m2-s are achieved, the steady state mode allows for the long term irradiation of components and systems in a fast neutron environment at a flux of up to 5×1015 n/m2-s. The SPR reactors are operated in a 9.2 meter diameter exposure cell, or Kiva, suitable for the irradiation of large test articles external to the reactors. Currently, a new upgraded version of SPR III (SPR HIM) is in fabrication; a unique feature of SPR HIM is its 190 mm (usable diameter) central irradiation cavity, the largest of any U.S. fast burst reactor. An improved cooling system permits continuous operation at power levels in excess of 20 kWt. The SPR Facility is also the operating site for a critical assembly which was used to characterize prototypic fuels in arrays appropriate for the Space Nuclear Thermal Propulsion Program. Work continues on use of the facility to design, build, and operate critical assemblies for a diverse customer base.

  1. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, D.W.; Schwint, A.E.; Hartwell, J.K.; Heber, E.M.; Trivillin, V.; Castillo, J.; Wentzeis, L.; Sloan, P.; Wemple, C.A.

    2004-10-04

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  2. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Amanda E. Schwint; John K. Hartwell; Elisa M. Heber; Veronica Trivillin; Jorge Castillo; Luis Wentzeis; Patrick Sloan; Charles A. Wemple

    2004-10-01

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  3. Reactor Neutrinos

    CERN Document Server

    Lasserre, T; Lasserre, Thierry; Sobel, Henry W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrino oscillation physics in the last years. It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the last undetermined neutrino mixing angle theta13. We conclude by opening on possible use of neutrinos for Society: NonProliferation of Nuclear materials and Geophysics.

  4. Experimental and MCNP5 based evaluation of neutron and gamma flux in the irradiation ports of the University of Utah research reactor

    Directory of Open Access Journals (Sweden)

    Noble Brooklyn

    2012-01-01

    Full Text Available Neutron and gamma flux environment of various irradiation ports in the University of Utah training, research, isotope production, general atomics reactor were experimentally assessed and fully modeled using the MCNP5 code. The experimental measurements were based on the cadmium ratio in the irradiation ports of the reactor, flux profiling using nickel wire, and gamma dose measurements using thermo luminescence dosimeter. Full 3-D MCNP5 reactor model was developed to obtain the neutron flux distributions of the entire reactor core and to compare it with the measured flux focusing at the irradiation ports. Integration of all these analysis provided the updated comprehensive neutron-gamma flux maps of the existing irradiation facilities of the University of Utah TRIGA reactor.

  5. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  6. Demolition to Green-Field conditions of the FRJ-1 (MERLIN) research reactor. Successes and hurdles in the demolition of a research reactor of the megawatt class; Der Rueckbau des Forschungsreaktors FRJ-1 (MERLIN) bis zur 'Gruenen Wiese'. Erfolge und Huerden beim Rueckbau eines Forschungsreaktors der Megawatt-Klasse

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, Burkhard; Printz, Rudolf; Matela, Karel; Zehbe, Carsten; Stauch, Bernhard; Zander, Iven [Forschungszentrum Juelich GmbH, Juelich (Germany)

    2010-02-15

    The Juelich-1 Research Reactor (FRJ-1), also referred to as MERLIN (Medium Energy Research Light Water Moderated Industrial Nuclear Reactor), was a light-water moderated and cooled swimming pool reactor of British design. The cornerstone in the erection of the reactor building was laid on June 11, 1958. Reactor operation was started on February 23, 1962. The plant was last run at a thermal power of 10 MW and shut down for good in 1985 after 23 years of operation. After the fuel elements had been removed and most of the experimental installations dismantled, some first steps towards demolition were taken in 1995. Demolition on a large scale began in 1996. September 8, 2008 was a special day: On the area of the former reactor hall, an oak tree was planted as a symbol of the 'green field' and of the original oak wood which had to make way for the construction of reactors in Juelich. An oak tree now stands in the place of the reactor unit. Was that all? It was not, for there were ancillary systems, operations, utility and hygiene buildings which had to be pulled down. Decontamination and clearance measurements were completed. The application for clearance was prepared and completed. Conventional demolition was started in 2009. After completion of that step, the last chapter about demolition of the FRJ-1 research reactor has been written, and the book can be closed. (orig.)

  7. Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jaluvka, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Wilson, E. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Foyto, L. P. [Univ. of Missouri, Columbia, MO (United States); Kutikkad, K. [Univ. of Missouri, Columbia, MO (United States); McKibben, J. C. [Univ. of Missouri, Columbia, MO (United States); Peters, N. J. [Univ. of Missouri, Columbia, MO (United States)

    2017-02-01

    This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members in the Research and Test Reactor Department at the Argonne National Laboratory (ANL) and the MURR Facility. MURR LEU conversion is part of an overall effort to develop and qualify high-density fuel within the U.S. High Performance Research Reactor Conversion (USHPRR) program conducted by the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization (M3).

  8. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, October--December 1976

    Energy Technology Data Exchange (ETDEWEB)

    Ferguson, J. B. [ed.

    1977-04-01

    Light water reactor safety research performed October through December 1976 is discussed. An analysis to determine the effect of emergency core coolant (ECC) injection location and pump speed on system response characteristics was performed. An analysis to evaluate the capability of commonly used critical heat flux (CHF) correlations to calculate the time of the first CHF in the Semiscale core during a loss-of-coolant experiment (LOCE) was performed. A test program and study to determine the effect thermocouples mounted on the outside fuel rod surfaces would have on the departure from nucleate boiling (DNB) phenomena in the LOFT core during steady state operation were completed. A correlation for use in predicting DNB heat fluxes in the LOFT core was developed. Tests of an experimental transit time flowmeter were completed. A nuclear test was performed to obtain fuel rod behavior data from four PWR-type rods during film boiling operation representative of PWR conditions. Preliminary results from the postirradiation examination of Test IE-1 fuel rods are given. Results of Irradiation Effects Tests IE-2 and IE-3 are given. Gap Conductance Test GC 2-1 was performed to evaluate the effects of fuel density, initial gap width, and fill gas composition on the pellet-cladding gap conductance.

  9. Carbon-14 in neutron-irradiated graphite for graphite-moderated reactors. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, Kimio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Matsuo, Hideto [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokyo (Japan)

    2002-12-01

    The graphite moderated gas cooled reactor operated by the Japan Atomic Power Company was stopped its commercial operation on March 1998, and the decommissioning process has been started. Graphite material is often used as the moderator and the reflector materials in the core of the gas cooled reactor. During the operation, a long life nuclide of {sup 14}C is generated in the graphite by several transmutation reactions. Separation of {sup 14}C isotope and the development of the separation method have been recognized to be critical issues for the decommissioning of the reactor core. To understand the current methodologies for the carbon isotope separation, literature on the subject was surveyed. Also, those on the physical and chemical behavior of {sup 14}C were surveyed. This is because the larger part of the nuclides in the graphite is produced from {sup 14}N by (n,p) reaction, and the location of them in the material tends to be different from those of the other carbon atoms. This report summarizes the result of survey on the open literature about the behavior of {sup 14}C and the separation methods, including the list of the literature on these subjects. (author)

  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. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  12. Fabrication of uranium dioxide fuel pellets in support of a SLOWPOKE-2 research reactor HEU to LEU core conversion

    Energy Technology Data Exchange (ETDEWEB)

    Bergeron, A. [Aomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2014-07-01

    The International Centre for Environmental and Nuclear Sciences (ICENS) at the University of the West Indies in Jamaica operates a SLOWPOKE-2 research reactor that is currently fuelled with highly-enriched uranium (HEU). As part of the Global Threat Reduction Initiative, Atomic Energy of Canada Ltd. has been subcontracted to fabricate low-enriched uranium (LEU) fuel for the ICENS SLOWPOKE-2. The low enriched uranium core consists of a fuel cage containing uranium dioxide fuelled elements. This paper describes the fabrication of the low-enriched uranium dioxide fuel pellets for the SLOWPOKE-2 core conversion. (author)

  13. Proceedings of the US Nuclear Regulatory Commission twentieth water reactor safety information meeting; Volume 2, Severe accident research, Thermal hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Weiss, A.J. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1993-03-01

    This three-volume report contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21--23, 1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchersfrom CEC, China, Finland, France, Germany, Japan, Spain and Taiwan. Selected papers have been processed separately for inclusion in the Energy Science and Technology Database.

  14. Measurement of in-phantom neutron flux and gamma dose in Tehran research reactor boron neutron capture therapy beam line.

    Science.gov (United States)

    Bavarnegin, Elham; Sadremomtaz, Alireza; Khalafi, Hossein; Kasesaz, Yaser

    2016-01-01

    Determination of in-phantom quality factors of Tehran research reactor (TRR) boron neutron capture therapy (BNCT) beam. The doses from thermal neutron reactions with 14N and 10B are calculated by kinetic energy released per unit mass approach, after measuring thermal neutron flux using neutron activation technique. Gamma dose is measured using TLD-700 dosimeter. Different dose components have been measured in a head phantom which has been designed and constructed for BNCT purpose in TRR. Different in-phantom beam quality factors have also been determined. This study demonstrates that the TRR BNCT beam line has potential for treatment of superficial tumors.

  15. Safety Evaluation Report Related to the Renewal of the Operating License for the TRIGA Training and Research Reactor at the University of Utah

    OpenAIRE

    United States Nuclear Regulatory Commission, Office of Nuclear Reactor Regulations

    1985-01-01

    This Safety Evaluation Report for the application filed by the University of Utah (UU) for a renewal of operating license R-126 to continue to operate a training and research reactor facility has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. The facility is owned and operated by the University of Utah and is located on its campus in Salt Lake City, Salt Lake County, Utah. The staff concludes that this training reactor facility can conti...

  16. Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core

    Energy Technology Data Exchange (ETDEWEB)

    PARMA JR.,EDWARD J.

    2000-01-01

    Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public.

  17. Diamond as a solid state micro-fission chamber for thermal neutron detection at the VR-1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pomorski, Michal; Mer-Calfati, Christine [CEA-LIST, Diamond Sensors Laboratory, 91191, Gif-sur-Yvette (France); Foulon, Francois [CEA, National Institute for Nuclear Science and Technology, 91191, Gif-sur-Yvette (France); Sklenka, Lubomir; Rataj, Jan; Bily, Tomas [Department of Nuclear Reactors,Faculty of Nuclear Science and Physical Engineering, Czech Technical University, V. Holesovickach 2, 180 00 PRAHA 8 (Czech Republic)

    2015-07-01

    Diamond exhibits a combination of properties which makes it attractive for neutron detection in hostile conditions. In the particular case of detection in a nuclear reactor, it is resilient to radiation, exhibits a natural low sensitivity to gamma rays, and its small size (as compared with that of gas ionisation chambers) enables fluency monitoring with a high position resolution. We report here on the use of synthetic CVD diamond as a solid state micro-fission chamber with U-235 converting material for in-core thermal neutron monitoring. Two types of thin diamond detectors were developed for this application. The first type of detector is fabricated using thin diamond membrane obtained by etching low-cost commercially available single crystal CVD intrinsic diamond, so called 'optical grade' material. Starting from a few hundred of micrometre thick samples, the sample is sliced with a laser and then plasma etched down to a few tenths of micrometre. Here we report the result obtained with a 17 μm thick device. The detection surface of this detector is equal to 1 mm{sup 2}. Detectors with surfaces up to 1 cm{sup 2} can be fabricated with this technique. The second type of detector is fabricated by growing successively two thin films of diamond, by the microwave enhanced chemical vapour deposition technique, on HPHT single crystal diamond. A first, a film of boron doped (p+) single crystal diamond, a few microns thick, is deposited. Then a second film of intrinsic diamond with a thickness of a few tens of microns is deposited. This results in a P doped, Intrinsic, Metal structure (PIM) structure in which the intrinsic volume id the active part of the detector. Here we report the results obtained with a 20 μm thick intrinsic whose detection surface is equal to 0.5 mm{sup 2}, with the possibility to enlarge the surface of the detector up to 1 cm{sup 2}. These two types of detector were tested at the VR-1 research reactor at the Czech Technical University in

  18. The present status of iodine chemistry research in Canada and its application to reactor safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, K.R. [Ontario Hydro Nuclear, Toronto (Canada); Kupferschmid, W.C.H.; Wren, J.C.; Ball, J.M. [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1996-12-01

    The current need to understand iodine chemistry in a reactor safety context has become more sharply focussed as the level of that understanding has advanced. At the same time, the situations of most concern within containment, from an iodine perspective, are also being redefined in the light of that understanding. The present paper summarises these developments. Over the past five years, considerable advances have occurred in our understanding of iodine chemistry under conditions of interest in reactor accidents. A number of key experiments have yielded important results in the areas of solution chemistry, the role of surfaces, the importance of organics and the effects of impurities. This understanding supplements the already substantial gains made in characterising the key roles of pH and the effects of radiation. All these factors underline the now evident fact that the kinetics of iodine are the controlling factor when radiation is involved, and that a number of reactive species, not present in thermal reactions, effectively control the observed volatility of iodine. In this paper, recent advances are summarised and the present status of our understanding of iodine chemistry is reviewed. Specifically, an attempt is made to identify those areas where our understanding appears to be relatively complete, and to flag the remaining critical areas where our attention is currently focussed. The state of our modelling capability is reviewed, as is the significance or related areas such as the role of mass transfer. Finally, an overview is presented of the significance of this work for reactor safety, and our expectations for its application over the near term future. (author) 2 figs., 12 refs.

  19. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

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

  20. Core map generation for the ITU TRIGA Mark II research reactor using Genetic Algorithm coupled with Monte Carlo method

    Energy Technology Data Exchange (ETDEWEB)

    Türkmen, Mehmet, E-mail: tm@hacettepe.edu.tr [Nuclear Engineering Department, Hacettepe University, Beytepe Campus, Ankara (Turkey); Çolak, Üner [Energy Institute, Istanbul Technical University, Ayazağa Campus, Maslak, Istanbul (Turkey); Ergün, Şule [Nuclear Engineering Department, Hacettepe University, Beytepe Campus, Ankara (Turkey)

    2015-12-15

    Highlights: • Optimum core maps were generated for the ITU TRIGA Mark II Research Reactor. • Calculations were performed using a Monte Carlo based reactor physics code, MCNP. • Single-Objective and Multi-Objective Genetic Algorithms were used for the optimization. • k{sub eff} and ppf{sub max} were considered as the optimization objectives. • The generated core maps were compared with the fresh core map. - Abstract: The main purpose of this study is to present the results of Core Map (CM) generation calculations for the İstanbul Technical University TRIGA Mark II Research Reactor by using Genetic Algorithms (GA) coupled with a Monte Carlo (MC) based-particle transport code. Optimization problems under consideration are: (i) maximization of the core excess reactivity (ρ{sub ex}) using Single-Objective GA when the burned fuel elements with no fresh fuel elements are used, (ii) maximization of the ρ{sub ex} and minimization of maximum power peaking factor (ppf{sub max}) using Multi-Objective GA when the burned fuels with fresh fuels are used. The results were obtained when all the control rods are fully withdrawn. ρ{sub ex} and ppf{sub max} values of the produced best CMs were provided. Core-averaged neutron spectrum, and variation of neutron fluxes with respect to radial distance were presented for the best CMs. The results show that it is possible to find an optimum CM with an excess reactivity of 1.17 when the burned fuels are used. In the case of a mix of burned fuels and fresh fuels, the best pattern has an excess reactivity of 1.19 with a maximum peaking factor of 1.4843. In addition, when compared with the fresh CM, the thermal fluxes of the generated CMs decrease by about 2% while change in the fast fluxes is about 1%.Classification: J. Core physics.

  1. Hydraulic modelling for analysis of the hot water layer stability in research reactor; Modelagem hidraulica para analise da estabilidade de camada de agua quente em reator de piscina

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, Rogerio; Yanagihara, Jurandir Itizo

    1995-12-31

    Pool reactors are research reactors, which allow easy access to the core and are simple to operate. Reactors of this kind operating at power levels higher than about one megawatt need a hot water layer at the surface of the pool, in order to keep surface activity below acceptable levels and enable free access to the upper part of the reactor. This work presents similitude criteria derived by dimensional analysis and by non dimensioning the basic equations to analyze this layer`s stability in a reduced scale model. The flow in the reactor is complex. It is impossible to consider all the phenomena with a single similitude criterion. The best would be to construct several models considering all the similitude criteria and then combine the results. Economical reasons and available time in the majority of the cases are a restrain to this procedure. Then, the most important criteria to the considered phenomenon must be chosen in order to give the best results. This work identifies three similitude criteria that were considered important to analyze the pool reactor`s hot water layer stability. (author) 16 refs., 3 figs.

  2. Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    El-Sahlamy, Neama M. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt)

    2017-11-15

    In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1{sub X}-A1, U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1{sub X}-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U{sub 3}Si{sub 2}-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

  3. Research Activities on Development of Piping Design Methodology of High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Nam-Su [Seoul National Univ. of Science and Technology, Seoul(Korea, Republic of); Won, Min-Gu [Sungkyukwan Univ., Suwon (Korea, Republic of); Oh, Young-Jin [KEPCO Engineering and Construction Co. Inc., Gimcheon (Korea, Republic of); Lee, Hyeog-Yeon; Kim, Yoo-Gon [Korea Atomic Energy Research Institute, Daejeon(Korea, Republic of)

    2016-10-15

    A SFR is operated at high temperature and low pressure compared with commercial pressurized water reactor (PWR), and such an operating condition leads to time-dependent damages such as creep rupture, excessive creep deformation, creep-fatigue interaction and creep crack growth. Thus, high temperature design and structural integrity assessment methodology should be developed considering such failure mechanisms. In terms of design of mechanical components of SFR, ASME B and PV Code, Sec. III, Div. 5 and RCC-MRx provide high temperature design and assessment procedures for nuclear structural components operated at high temperature, and a Leak-Before-Break (LBB) assessment procedure for high temperature piping is also provided in RCC-MRx, A16. Three web-based evaluation programs based on the current high temperature codes were developed for structural components of high temperature reactors. Moreover, for the detailed LBB analyses of high temperature piping, new engineering methods for predicting creep C*-integral and creep COD rate based either on GE/EPRI or on reference stress concepts were proposed. Finally, the numerical methods based on Garofalo's model and RCC-MRx have been developed, and they have been implemented into ABAQUS. The predictions based on both models were compared with the experimental results, and it has been revealed that the predictions from Garafalo's model gave somewhat successful results to describe the deformation behavior of Gr. 91 at elevated temperatures.

  4. The DRAGON aerosol research facility to study aerosol behaviour for reactor safety applications

    Energy Technology Data Exchange (ETDEWEB)

    Suckow, Detlef; Guentay, Salih [Paul Scherrer Institut, Villigen (Switzerland)

    2008-07-01

    During a severe accident in a nuclear power plant fission products are expected to be released in form of aerosol particles and droplets. To study the behaviour of safety relevant reactor components under aerosol loads and prototypical severe accident conditions the multi-purpose aerosol generation facility DRAGON is used since 1994 for several projects. DRAGON can generate aerosol particles by the evaporation-condensation technique using a plasma torch system, fluidized bed and atomization of particles suspended in a liquid. Soluble, hygroscopic aerosol (i.e. CsOH) and insoluble aerosol particles (i.e. SnO{sub 2}, TiO{sub 2}) or mixtures of them can be used. DRAGON uses state-of-the-art thermal-hydraulic, data acquisition and aerosol measurement techniques and is mainly composed of a mixing chamber, the plasma torch system, a steam generator, nitrogen gas and compressed air delivery systems, several aerosol delivery piping, gas heaters and several auxiliary systems to provide vacuum, coolant and off-gas treatment. The facility can be operated at system pressure of 5 bars, temperatures of 300 deg. C, flow rates of non-condensable gas of 900 kg/h and steam of 270 kg/h, respectively. A test section under investigation is attached to DRAGON. The paper summarizes and demonstrates with the help of two project examples the capabilities of DRAGON for reactor safety studies. (authors)

  5. Validation Calculations for the Application of MARS Code to the Safety Analysis of Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Kim, H.; Chae, H. T.; Lim, I. C

    2006-10-15

    In order to investigate the applicability of MARS code to the accident analysis of the HANARO and other RRs, the following test data were simulated. Test data of the HANARO design and operation, Test data of flow instability and void fraction from published documents, IAEA RR transient data in TECDOC-643, Brazilian IEA-R1 experimental data. For the simulation of the HANARO data with finned rod type fuels at low pressure and low temperature conditions, MARS code, developed for the transient analysis of power reactors, was modified. Its prediction capability was assessed against the experimental data for the HANARO. From the assessment results, it can be said that the modified MARS code could be used for analyzing the thermal hydraulic transient of the HANARO. Some other simulations such as flow instability test and reactor transients were also done for the application of MARS code to RRs with plate type fuels. In the simulation for these cases, no modification was made. The results of simulated cases show that the MARS code can be used to the transient analysis of RRs with careful considerations. In particular, it seems that an improvement on a void model may be necessary for dealing with the phenomena in high void conditions.

  6. IAEA coordinated research program on `harmonization and validation of fast reactor thermomechanical and thermohydraulic codes using experimental data`. 1. Thermohydraulic benchmark analysis on high-cycle thermal fatigue events occurred at French fast breeder reactor Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-06-01

    A benchmark exercise on `Tee junction of Liquid Metal Fast Reactor (LMFR) secondary circuit` was proposed by France in the scope of the said Coordinated Research Program (CRP) via International Atomic Energy Agency (IAEA). The physical phenomenon chosen here deals with the mixture of two flows of different temperature. In a LMFR, several areas of the reactor are submitted to this problem. They are often difficult to design, because of the complexity of the phenomena involved. This is one of the major problems of the LMFRs. This problem has been encountered in the Phenix reactor on the secondary loop, where defects in a tee junction zone were detected during a campaign of inspections after an operation of 90,000 hours of the reactor. The present benchmark is based on an industrial problem and deal with thermal striping phenomena. Problems on pipes induced by thermal striping phenomena have been observed in some reactors and experimental facilities coolant circuits. This report presents numerical results on thermohydraulic characteristics of the benchmark problem, carried out using a direct numerical simulation code DINUS-3 and a boundary element code BEMSET. From the analysis with both the codes, it was confirmed that the hot sodium from the small pipe rise into the cold sodium of the main pipe with thermally instabilities. Furthermore, it was indicated that the coolant mixing region including the instabilities agrees approximately with the result by eye inspections. (author)

  7. Low enriched uranium foil plate target for the production of fission Molybdenum-99 in Pakistan Research Reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Mushtaq, A. [Isotope Production Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad, Federal Area 44000 (Pakistan)], E-mail: mushtaqa@pinstech.org.pk; Iqbal, Masood; Bokhari, Ishtiaq Hussain; Mahmood, Tayyab [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad, Federal Area 44000 (Pakistan)

    2009-04-15

    Low enriched uranium foil (19.99% {sup 235}U) will be used as target material for the production of fission Molybdenum-99 in Pakistan Research Reactor-1 (PARR-1). LEU foil plate target proposed by University of Missouri Research Reactor (MURR) will be irradiated in PARR-1 for the production of 100Ci of Molybdenum-99 at the end of irradiation, which will be sufficient to prepare required {sup 99}Mo/{sup 99m}Tc generators at Pakistan Institute of Nuclear Science and Technology, Islamabad (PINSTECH) and its supply in the country. Neutronic and thermal hydraulic analysis for the fission Molybdenum-99 production at PARR-1 has been performed. Power levels in target foil plates and their corresponding irradiation time durations were initially determined by neutronic analysis to have the required neutron fluence. Finally, the thermal hydraulic analysis has been carried out for the proposed design of the target holder using LEU foil plates for fission Molybdenum-99 production at PARR-1. Data shows that LEU foil plate targets can be safely irradiated in PARR-1 for production of desired amount of fission Molybdenum-99.

  8. 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.

  9. The Implementation of an Integrated Management System for TRIGA Research Reactor at LENA (Laboratory of Applied Nuclear Energy) - University of Pavia (Italy) -

    Energy Technology Data Exchange (ETDEWEB)

    Cagnazzo, M.; Tigliole, A. Borio Di; Magrotti, G.; Manera, S.; Marchetti, F.; Prata, M.; Salvini, A. [Laboratory of Applied Nuclear Energy (LENA), University of Pavia (Italy); Giordano, M. [Innovation and Management Systems Division, University of Pavia (Italy); Boogaard, J.P.; Bradley, E.; Vincze, P. [International Atomic Energy Agency (IAEA), Vienna (Austria)

    2011-07-01

    The Laboratory of Applied Nuclear Energy ('LENA') is an Interdepartmental Research Centre of the University of Pavia which operate, among other facilities, a 250 kW TRIGA Mark II Research Nuclear Reactor. The reactor is at the disposal of researchers from Pavia University and of other users, both public and private, for research activities, training and education and other services. The Centre itself carries out research and training activities and provides services for private enterprises, encouraging the transfer of the results of nuclear technology research to the production system, including the education and training of specialists in nuclear technology. The prerequisite for the management of the reactor is the satisfaction of all stakeholders requirements, among which safety constraints, efficiency and effectiveness in the delivery of the services. In order to continuously improve the safety and quality of reactor management and the accomplishment of the stakeholder requirements, LENA decided to implement an Integrated Management System in accordance with International Standard ISO 9001:2008. This choice allowed to satisfy both national and international compulsory requirements (i.e. safe reactor operation and maintenance) and typical ISO 9001 requirements (as e.g. continuous improvement, users/stakeholders care and satisfaction). In addition, through this systematic and graded approach, that led to the standardization of all processes involved in reactor operation and maintenance, all the aspects of the reactor management mentioned in the IAEA publication The Management System for Facilities and Activities (IAEA Safety Standards Series No. GS-R-3) were also satisfied. This publication, in facts, provides a guidance for establishing, implementing, assessing and continually improving a management system for facilities and activities that integrates safety, health, environmental, security, quality and economic elements. (author)

  10. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, October--December 1975

    Energy Technology Data Exchange (ETDEWEB)

    1976-05-01

    Light water reactor safety activities performed during October--December 1975 are reported. The blowdown heat transfer tests series of the Semiscale Mod-1 test program was completed. In the LOFT Program, preparations were made for nonnuclear testing. The Thermal Fuels Behavior Program completed a power-cooling-mismatch test and an irradiation effects test on PWR-type fuel rods. Model development and verification efforts of the Reactor Behavior Program included developing new analysis models for the RELAP4 computer code, subroutines for the FRAP-S and FRAP-T codes, and new models for predicting reactor fuel restructuring and zircaloy cladding behavior; an analysis of post-CHF fuel behavior was made using FRAP-T.

  11. Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. A. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Wilson, E. H. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Foyto, L. P. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Kutikkad, K. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; McKibben, J. C. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Peters, N. J. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Cowherd, W. M. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program; Rickman, B. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program

    2014-12-01

    This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo

  12. Quarterly technical report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, July--September 1975

    Energy Technology Data Exchange (ETDEWEB)

    1976-02-01

    Light water reactor safety activities performed during July through September 1975 are summarized. The isothermal blowdown test series of the Semiscale Mod-1 test program has provided data for evaluation of break flow phenomena and analyses of piping flow regimes and pump performance. In the LOFT Program, measurement uncertainties were evaluated. The Thermal Fuels Behavior Program completed two power-cooling-mismatch tests on PWR-type fuel rods to investigate critical heat flux characteristics. Model development and verification efforts of the Reactor Behavior Program included development of the SPLEN1 computer code, subroutines for the FRAP-T code, verification of RELAP4, and results of the Halden Recycle Plutonium Experiment.

  13. Automated system for neutron activation analysis determination of short lived isotopes at The DOW Chemical Company's TRIGA research reactor

    Science.gov (United States)

    Zieman, J. J.; Rigot, W. L.; Romick, J. D.; Quinn, T. J.; Kocher, C. W.

    1994-12-01

    An automated neutron activation analysis (NAA) system for the determination of short lived isotopes was constructed at The DOW Chemical Company's TRIGA Research Reactor in 1993. The NAA group of the Analytical Sciences Laboratory uses the reactor for thousands of analyses each year and therefore automation is important to achieve and maintain high throughput and precision (productivity). This project is complementary to automation of the long-lived counting facilities (see Romick et al., these Proceedings). Canberra/Nuclear Data Systems DEC-based software and electronics modules and an I/O mounting board are the basic commercial components. A Fortran program on a VAX computer controls I/O via ethernet to an Acquisition Interface Module (AIM). The AIM controls the γ spectrometer modules and is interfaced to a Remote Parallel Interface (RPI) module which controls the pneumatic transfer apparatus with TTL signals to the I/O mounting board. Near-infrared sensors are used to monitor key points in the transfer system. Spectra are acquired by a single HPGe detector mounted on a sliding rail to allow flexible and more reproducible counting geometries than with manual sample handling. The maximum sample size is 8 ml in a heat-sealed two dram vial. The sample vial is nested into a "rabbit" vial for irradiation which can be automatically removed prior to spectrum collection. The system was designed to be used by the reactor operator at the control console without the aid of an additional experimenter. Applications include the determination of selenium and silver in coal and water, fluorine in tetra-fluoro ethylene (TFE) coated membranes, aluminum and titanium in composite materials and trace fluorine in non-chlorinated cleaning solvents. Variable dead time software allows analysis for 77mSe despite high dead times from 16N encountered in samples.

  14. Research of lithium capillary-pore systems for fusion reactor plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Evtikhin, V.A. E-mail: evtikhin@protein.bio.msu.ru; Vertkov, A.V.; Lyublinski, I.E.; Khripunov, B.I.; Petrov, V.B.; Mirnov, S.V

    2002-12-01

    To date there is no adequate solution for high heat load plasma facing components of the next step fusion reactor among solid material options. A lithium-filled capillary porous systems (CPS) was proposed as a plasma facing material and experimental work on this subject is now in progress. Steady-state experiments with CPS-based target and lithium supply systems have shown successful operation at heat fluxes of 1-10 MW/m{sup 2} during several hours. Experimental data is obtained on lithium CPS stability at heat flux up to 25-50 MW/m{sup 2}. The lithium CPS behaviour in contact with real tokamak plasma is considered for normal discharge condition at 10 MW/m{sup 2} and for plasma disruption at 15 MJ/m{sup 2}. Erosion mechanism of lithium under tokamak plasma impact was analysed. Stability of lithium CPS in tokamak conditions was shown.

  15. Study of Reduced-Enrichment Uranium Fuel Possibility for Research Reactors

    Directory of Open Access Journals (Sweden)

    Ruppel V.A.

    2015-01-01

    Full Text Available Having analyzed the results obtained in the work, it is possible to conclude that the flux density of fast and thermal neutrons in the shell of fuel elements in EFA in REU-zone decreased on average by 5% for UO2 fuel and by 7% for U9%Mo fuel. Change of neutrons flux density during the cycle does not exceed 4% for both fuel types. On average the fuel burnup in reactor core during the cycle for UO2 and U9%Mo increased by 2.8%. It is 1% less that in HEU-zone, which is conditioned by higher initial loading of 235U in fuel assembly with REU fuel.

  16. Natural and man-made radioactivity in soils and plants around the research reactor of Inshass.

    Science.gov (United States)

    Higgy, R H; Pimpl, M

    1998-12-01

    The specific radioactivities of the U-series, 232Th, 137Cs and 40K were measured in soil samples around the Inshass reactor in Cairo, using a gamma-ray spectrometer with a HpGe detector. The alpha activity of 238U, 234U and 235U was measured in the same soil samples by surface barrier detectors after radiochemical separation and the obtained results were compared with the specific activities determined by gamma-measurements. The alpha-activity of 238Pu, 239+240Pu, 241Am, 242Cm and 244Cm was measured after radiochemical separation by surface barrier detectors for both soil and plant samples. Then beta-activity of 241Pu was measured using liquid scintillation spectrometry.

  17. Study on the fuel cycle cost of gas turbine high temperature reactor (GTHTR300). Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Takei, Masanobu; Katanishi, Shoji; Nakata, Tetsuo; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Oda, Takefumi; Izumiya, Toru [Nuclear Fuel Industries, Ltd., Tokyo (Japan)

    2002-11-01

    In the basic design of gas turbine high temperature reactor (GTHTR300), reduction of the fuel cycle cost has a large benefit of improving overall plant economy. Then, fuel cycle cost was evaluated for GTHTR300. First, of fuel fabrication for high-temperature gas cooled reactor, since there was no actual experience with a commercial scale, a preliminary design for a fuel fabrication plant with annual processing of 7.7 ton-U sufficient four GTHTR300 was performed, and fuel fabrication cost was evaluated. Second, fuel cycle cost was evaluated based on the equilibrium cycle of GTHTR300. The factors which were considered in this cost evaluation include uranium price, conversion, enrichment, fabrication, storage of spent fuel, reprocessing, and waste disposal. The fuel cycle cost of GTHTR300 was estimated at about 1.07 yen/kWh. If the back-end cost of reprocessing and waste disposal is included and assumed to be nearly equivalent to LWR, the fuel cycle cost of GTHTR300 was estimated to be about 1.31 yen/kWh. Furthermore, the effects on fuel fabrication cost by such of fuel specification parameters as enrichment, the number of fuel types, and the layer thickness were considered. Even if the enrichment varies from 10 to 20%, the number of fuel types change from 1 to 4, the 1st layer thickness of fuel changes by 30 {mu}m, or the 2nd layer to the 4th layer thickness of fuel changes by 10 {mu}m, the impact on fuel fabrication cost was evaluated to be negligible. (author)

  18. Study on Reduced-Moderation Water Reactor (RMWR) core design. Joint research report (FY1998-1999)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-09-01

    The Reduce-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor aiming at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. Japan Atomic Energy Research Institute (JAERI) started a joint research program for conceptual design of RMWR core in collaboration with the Japan Atomic Power Company (JAPC) since 1998. The research area includes the RMWR core conceptual designs, development of analysis methods for rector physics and thermal-hydraulics to design the RMWR cores with higher accuracy and preparation of MOX critical experiment to confirm the feasibility from the reactor physics point of view. The present report describes the results of joint research program 'RMWR core design Phase 1' performed by JAERI and JAPC in FY 1998 and 1999. The results obtained from the joint research program are as follows: Conceptual design study on the RMWR core has been performed. A core concept with a conversion ratio more than about 1 is basically feasible to multiple recycling of plutonium. Investigating core characteristics at the equilibrium, some promising core concepts to satisfy above aims have been established. As for BWR-type concepts with negative void reactivity coefficients, three types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.1, (2) one feasible to attain operation cycle of about 2 years and burn-up of about 60 GWd/t with conversion ratio more than 1 or (3) one in simple design based on the ABWR assembly and without blanket attaining conversion ratio more than 1. And as for PWR-type concepts with negative void reactivity coefficients, two types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.05 by using heavy water as a coolant and (2) one feasible to attain conversion ratio about l by using light water. In the study of nuclear calculation method, a reactor analysis code

  19. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  20. 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...

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

    Science.gov (United States)

    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.; Nörtershäuser, W.; Repp, J.; Smorra, C.; Trautmann, N.; Weber, C.

    2008-09-01

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

  2. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  3. Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for kijang research reactor

    Directory of Open Access Journals (Sweden)

    Jong Man Park

    2017-08-01

    Full Text Available The construction project of the Kijang research reactor (KJRR, which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm3, was selected to achieve higher fuel efficiency and performance than are possible when using U3Si2/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm3, were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

  4. Analysis on the post-irradiation examination of the HANARO miniplate-1 irradiation test for Kijang research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Man; Tahk, Young Wook; Jeong, Yong Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); and others

    2017-08-15

    The construction project of the Kijang research reactor (KJRR), which is the second research reactor in Korea, has been launched. The KJRR was designed to use, for the first time, U–Mo fuel. Plate-type U–7 wt.% Mo/Al–5 wt.% Si, referred to as U–7Mo/Al–5Si, dispersion fuel with a uranium loading of 8.0 gU/cm{sup 3}, was selected to achieve higher fuel efficiency and performance than are possible when using U{sub 3}Si{sub 2}/Al dispersion fuel. To qualify the U–Mo fuel in terms of plate geometry, the first miniplates [HANARO Miniplate (HAMP-1)], containing U–7Mo/Al–5Si dispersion fuel (8 gU/cm{sup 3}), were fabricated at the Korea Atomic Energy Research Institute and recently irradiated at HANARO. The PIE (Post-irradiation Examination) results of the HAMP-1 irradiation test were analyzed in depth in order to verify the safe in-pile performance of the U–7Mo/Al–5Si dispersion fuel under the KJRR irradiation conditions. Nondestructive analyses included visual inspection, gamma spectrometric mapping, and two-dimensional measurements of the plate thickness and oxide thickness. Destructive PIE work was also carried out, focusing on characterization of the microstructural behavior using optical microscopy and scanning electron microscopy. Electron probe microanalysis was also used to measure the elemental concentrations in the interaction layer formed between the U–Mo kernels and the matrix. A blistering threshold test and a bending test were performed on the irradiated HAMP-1 miniplates that were saved from the destructive tests. Swelling evaluation of the U–Mo fuel was also conducted using two methods: plate thickness measurement and meat thickness measurement.

  5. Thermo-fluid analysis of water cooled research reactors in natural convection; Analise termofluidodinamica de reatores nucleares de pesquisa refrigerados a agua em regime de conveccao natural

    Energy Technology Data Exchange (ETDEWEB)

    Veloso, Maria Auxiliadora Fortini

    2004-07-01

    The STHIRP-1 computer program, which fundamentals are described in this work, uses the principles of the subchannels analysis and has the capacity to simulate, under steady state and transient conditions, the thermal and hydraulic phenomena which occur inside the core of a water-refrigerated research reactor under a natural convection regime. The models and empirical correlations necessary to describe the flow phenomena which can not be described by theoretical relations were selected according to the characteristics of the reactor operation. Although the primary objective is the calculation of research reactors, the formulation used to describe the fluid flow and the thermal conduction in the heater elements is sufficiently generalized to extend the use of the program for applications in power reactors and other thermal systems with the same features represented by the program formulations. To demonstrate the analytical capacity of STHIRP-l, there were made comparisons between the results calculated and measured in the research reactor TRIGA IPR-R1 of CDTN/CNEN. The comparisons indicate that the program reproduces the experimental data with good precision. Nevertheless, in the future there must be used more consistent experimental data to corroborate the validation of the program. (author)

  6. Recent Research of Thorium Molten-Salt Reactor from a Sustainability Viewpoint

    Directory of Open Access Journals (Sweden)

    Takashi Kamei

    2012-09-01

    Full Text Available The most important target of the concept “sustainability” is to achieve fairness between generations. Its expanding interpolation leads to achieve fairness within a generation. Thus, it is necessary to discuss the role of nuclear power from the viewpoint of this definition. The history of nuclear power has been the control of the nuclear fission reaction. Once this is obtained, then the economy of the system is required. On the other hand, it is also necessary to consider the internalization of the external diseconomy to avoid damage to human society caused by the economic activity itself, due to its limited capacity. An extreme example is waste. Thus, reducing radioactive waste resulting from nuclear power is essential. Nuclear non-proliferation must be guaranteed. Moreover, the FUKUSHIMA accident revealed that it is still not enough that human beings control nuclear reaction. Further, the most essential issue for sustaining use of one technology is human resources in manufacturing, operation, policy-making and education. Nuclear power will be able to satisfy the requirements of sustainability only when these subjects are addressed. The author will review recent activities of a thorium molten-salt reactor (MSR as a cornerstone for a sustainable society and describe its objectives and forecasts.

  7. Characteristic differences of LEU and HEU cores at the German FRJ-2 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nabbi, R.; Wolters, J.; Damm, G. [Central Research Reactor Division, Forschungszentrum Juelich, 52425 Juelich (Germany)

    2002-07-01

    As a sophisticated computational method for reactor physics analysis and fuel management an MCNP model in very high fidelity was developed and coupled with a depletion code and applied to the HEU-LEU core conversion study. The analysis show that as a consequence of the high amount of U-238, the amount of U-235 in the LEU core is about 14% higher than in the HEU core. The reduction of the thermal flux varies between 16% (core) and 5% in the reflector zone. The rate of U-235 burnup in the LEU core is approx. 11.5% lower which allows an extension of irradiation time. Due to the effect of neutron spectrum the worth of the absorber system decreases in an LEU core by 17% resulting in a decrease of shutdown and excess reactivity. The kinetic parameters of the core are slightly reduced causing changes in the reactivity values and transient behavior of the core. The moderator coefficient is decreased by 18% and the Doppler coefficient is increased by 63%. Due to shortening of the absorption length of the fission neutrons the prompt neutron lifetime is reduced by 7%. (author)

  8. NRC review of Electric Power Research Institute's Advanced Light Reactor Utility Requirements Document - Program summary, Project No. 669

    Energy Technology Data Exchange (ETDEWEB)

    1992-08-01

    The staff of the US Nuclear Regulatory Commission has prepared Volume 1 of a safety evaluation report (SER), NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Program Summary,'' to document the results of its review of the Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document.'' This SER provides a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  9. Releasable activity and maximum permissible leakage rate within a transport cask of Tehran Research Reactor fuel samples

    Directory of Open Access Journals (Sweden)

    Rezaeian Mahdi

    2015-01-01

    Full Text Available Containment of a transport cask during both normal and accident conditions is important to the health and safety of the public and of the operators. Based on IAEA regulations, releasable activity and maximum permissible volumetric leakage rate within the cask containing fuel samples of Tehran Research Reactor enclosed in an irradiated capsule are calculated. The contributions to the total activity from the four sources of gas, volatile, fines, and corrosion products are treated separately. These calculations are necessary to identify an appropriate leak test that must be performed on the cask and the results can be utilized as the source term for dose evaluation in the safety assessment of the cask.

  10. Measuring Neutron Spectrum at MIT Research Reactor Utilizing He-3 Bonner Cylinder Approach with an Unfolding Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Leder, A. [MIT; Anderson, A. J. [Chicago U., KICP; Billard, J. [Lyon, IPN; Figueroa-Feliciano, E. [Northwestern U.; Formaggio, J. A. [MIT; Hasselkus, C. [Wisconsin U., Madison; Newman, E. [MIT; Palladino, K. [Wisconsin U., Madison; Phuthi, M. [MIT; Winslow, L. [MIT; Zhang, L. [MIT

    2017-10-02

    The Ricochet experiment seeks to measure Coherent (neutral-current) Elastic Neutrino-Nucleus Scattering using dark-matter-style detectors with sub-keV thresholds placed near a neutrino source, such as the MIT (research) Reactor (MITR), which operates at 5.5 MW generating approximately 2.2e18 neutrinos/second at the core. Currently, Ricochet is characterizing the backgrounds at MITR, the main component of which comes in the form of neutrons emitted from the core simultaneous with the neutrino signal. To characterize this background, we wrapped a Bonner cylinder around a He-3 thermal neutron detector, whose data was then unfolded to produce a neutron energy spectrum across several orders of magnitude. We discuss the resulting spectrum and its implications for deploying Ricochet in the future at the MITR site as well as the feasibility of reducing this background level via the addition of polyethylene shielding around the detector setup.

  11. Neutronic and thermal-hydraulic analysis of fission molybdenum-99 production at Tehran Research Reactor using LEU plate targets.

    Science.gov (United States)

    Abedi, Ebrahim; Ebrahimkhani, Marzieh; Davari, Amin; Mirvakili, Seyed Mohammad; Tabasi, Mohsen; Maragheh, Mohammad Ghannadi

    2016-12-01

    Efficient and safe production of molybdenum-99 ( 99 Mo) radiopharmaceutical at Tehran Research Reactor (TRR) via fission of LEU targets is studied. Neutronic calculations are performed to evaluate produced 99 Mo activity, core neutronic safety parameters and also the power deposition values in target plates during a 7 days irradiation interval. Thermal-hydraulic analysis has been also carried out to obtain thermal behavior of these plates. Using Thermal-hydraulic analysis, it can be concluded that the safety parameters are satisfied in the current study. Consequently, the present neutronic and thermal-hydraulic calculations show efficient 99 Mo production is accessible at significant activity values in TRR current core configuration. Copyright © 2016 Elsevier Ltd. All rights reserved.

  12. Derived release limits for the greek research reactor site based on a diagnostic atmospheric modeling system for irregular terrain.

    Science.gov (United States)

    Varvayanni, M; Catsaros, N; Antonopoulos-Domis, M

    2005-04-01

    The upper limits for the rate of release of radionuclides into the atmosphere, i.e., the "derived release limits," are calculated for the Greek Research Reactor (GRR-1) in order to determine possible operational schemes compatible with the effective dose limits for the general population. GRR-1 is located at the northwestern foot of Hymettos Mountain and at the eastern border of the urbanized area of Athens basin. Due to the topographic complexity of the region, the meteorological and atmospheric dispersion calculations were based on a numerical modeling system that is especially designed to work over irregular terrains by using a prismatic unstructured grid. The calculation of derived release limits was made using guidelines and methods that conform to the system of dose limits prescribed by the European radiation protection regulations.

  13. The awareness of the functional and near population with the relation to the research nuclear reactor IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

    Vanni, Silvia R.; Martins, Maria da Penha S. [Centro Tecnologico da Marinha (CTMSP), SP (Brazil); Sabundjian, Gaiane, E-mail: gdjian@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    After the natural accident that hit Japan in the beginning of March of 2011, and that ended into an accident of great proportions in the nuclear installations of Fukushima, it has now the debate over the lack of information that the population in general has over the nuclear energy. The dissemination of information, about the operation and security of the nuclear reactors, has the purpose of softening the effect that the pessimistic atmosphere has over its using. This study was reinforced by the memories of serious consequences due to other nuclear accidents that have already happened (Chernobyl, Three-Mile and Hiroshima/Nagasaki event), bringing insecurity, fear and even revenge from part of the public. Over all, people are not sufficiently informed about the positives and negatives aspects of the nuclear energy. It is necessary the adoption of a clear and aware policy with the population, about the pacific use of nuclear energy. Today, the international and national organizations of control of nuclear energy, the International Atomic Energy Agency (IAEA) and the Comissao Nacional de Energia Nuclear (CNEN), have respectively, published information about this subject using a more professional way and of hard access for the public in general. This work has the goal of checking the level of information that the population of workers and individuals of the close public to the research nuclear reactor IEA-R1, located in the Institute of Nuclear Research (IPEN), University City, Sao Paulo, Brazil, has over it. The way used for this study, involved questionnaires with straight questions and of simple language over the subject, to people of all different social, economic and cultural classes, from 12 to 80 years old. From the results found after this work, it was verified the necessity to elaborate a project of awareness of information and clarification about the nuclear energy, using ways of communication that exist and that are easy for the public to understand. (author)

  14. 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.

  15. Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

    Science.gov (United States)

    Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Unruh, Troy C.; McGregor, Douglas S.; Roberts, Jeremy A.

    2017-08-01

    Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

  16. Evaluation of operational incidents in the research reactor RP-10 according to scale INES; Evaluacion de incidentes operacionales en el reactor de investigacion RP-10 segun escala INES

    Energy Technology Data Exchange (ETDEWEB)

    Arrieta, Rolando W.B.; Vela Mora, Mariano, E-mail: rarrieta@ipen.gob.pe, E-mail: mvela@ipen.gob.pe [Instituto Peruano de energia Nuclear, Lima (Peru). Dept. de Operacion de Reactores

    2013-07-01

    This report presents the evaluation of the events in 2011 in the RP-10 Nuclear Reactor Nuclear Center Huarangal from the point of view of safety. To classify these events produced is used Scale International Nuclear and Radiological Event Scale (INES) to facilitate a common understanding between the technical community, the media and the general public. From the results we can say that in 2011 all related to security events that occurred in the RP -10 are classified as 'below scale' or no safety significance. (author)

  17. Research on the improvement of nuclear safety -Thermal hydraulic tests for reactor safety system-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Moon Kee; Park, Choon Kyung; Yang, Sun Kyoo; Chun, Se Yung; Song, Chul Hwa; Jun, Hyung Kil; Jung, Heung Joon; Won, Soon Yun; Cho, Yung Roh; Min, Kyung Hoh; Jung, Jang Hwan; Jang, Suk Kyoo; Kim, Bok Deuk; Kim, Wooi Kyung; Huh, Jin; Kim, Sook Kwan; Moon, Sang Kee; Lee, Sang Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-06-01

    The present research aims at the development of the thermal hydraulic verification test technology for the safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. In this research, test facilities simulating the primary coolant system and safety system are being constructed for the design verification tests of the existing and advanced nuclear power plant. 97 figs, 14 tabs, 65 refs. (Author).

  18. Development of a digital card to simulate period transients in research reactors; Desenvolvimento de um cartao digital para simulacao da variacao do periodo em reatores de pesquisa

    Energy Technology Data Exchange (ETDEWEB)

    Masotti, Paulo Henrique Ferraz

    1999-07-01

    This work presents the development of a card to be used in a 'slot' of a micro-computer for evaluation of a nuclear channel used to monitor the start up of nuclear reactors. The results of the bench tests showed good linearity and 2% error deviation in the entire range of operation. Fields tests, performed with the start up channel of IEA-R1 research reactor showed that the card is an excellent device to verify the performance of the channel during steady state, and transient conditions. (author)

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

    Directory of Open Access Journals (Sweden)

    Maritza Rodríguez - Gual

    2010-05-01

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

  20. Seismic Safety Margins Research Programs. Assessment of potential increases in risk due to degradation of steam generator and reactor coolant pump supports. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bohn, M. P.; Wells, J. E.; Shieh, L. C.; Cover, L. E.; Streit, R. L.

    1983-08-01

    During the NRC licensing review for the North Anna Units 1 and 2 pressurized-water reactors (PWRs), questions were raised regarding the potential for low-fracture toughness of steam-generator and reactor-coolant-pump supports. Because other PWRs may face similar problems, this issue was incorporated into the NRC Program for Resolution of Generic Issues. The work described in this report was performed to provide the NRC with a quantitative evaluation of the value/impact implications of the various options of resolving the fracture-toughness question. This report presents an assessment of the probabilistic risk associated with nil-ductility failures of steam-generator and reactor-coolant-pump structural-support systems during seismic events, performed using the Seismic Safety Margins Research Program codes and data bases.

  1. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  2. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  3. Current status of neutron activation analysis using the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen Van Suc; Nguyen Mong Sinh [Nuclear Research Institute, Dalat (Viet Nam)

    1999-10-01

    Neutron activation analysis is one of the most sensitive, rapid, accurated methods for determination of trace elements in different materials. A review is made of the current status of the activities and the results in studying and developing NAA (Neutron Activation Analysis) at the Dalat Nuclear Research Institute and applying this method to different sectors of science and technology in Vietnam. (author)

  4. Neutronic simulation of a research reactor core of ( Th, U)O2 fuel ...

    Indian Academy of Sciences (India)

    2013-01-08

    Jan 8, 2013 ... 3European Organization for Nuclear Research, CERN, 1211 Geneva 23, Switzerland. 4Department of Energy .... ENDF/B S(α, β) scattering law and (2) elastic scattering with no change in the out- going neutron ..... of zirlo and other zirconium alloys, International Nuclear Atlantic Conference – INAC (2007).

  5. Research on friction coefficient of nuclear Reactor Vessel Internals Hold Down Spring: Stress coefficient test analysis method

    Energy Technology Data Exchange (ETDEWEB)

    Linjun, Xie, E-mail: linjunx@zjut.edu.cn [College of Mechanical Engineering, Zhejiang University of Technology, Hangzhou 310014 (China); Guohong, Xue; Ming, Zhang [Shanghai Nuclear Engineering Research & Design Institute, Shanghai 200233 (China)

    2016-08-01

    Graphical abstract: HDS stress coefficient test apparatus. - Highlights: • This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. • The mathematical relation between the load and the strain is obtained about the HDS, and the mathematical model of the stress coefficient and the friction coefficient is established. So, a set of test apparatuses for obtaining the stress coefficient is designed according to the model scaling criterion and the friction coefficient of the K1000 HDS is calculated to be 0.336 through the obtained stress coefficient. • The relation curve between the theoretical load and the friction coefficient is obtained through analysis and indicates that the change of the friction coefficient f would influence the pretightening load under the condition of designed stress. The necessary pretightening load in the design process is calculated to be 5469 kN according to the obtained friction coefficient. Therefore, the friction coefficient and the pretightening load under the design conditions can provide accurate pretightening data for the analysis and design of the reactor HDS according to the operations. - Abstract: This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. By carrying out tests and researches through a stress testing technique, P–σ curves in loading and unloading processes of the HDS are obtained and the stress coefficient k{sub f} of the HDS is obtained. So, the

  6. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [Brookhaven National Lab., Upton, NY (United States)] [comp.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

  7. Effect of kinetic parameters on simultaneous ramp reactivity insertion plus beam tube flooding accident in a typical low enriched U{sub 3}Si{sub 2}-Al fuel-based material testing reactor-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nasir, Rubina; Mirza, Nasir M. [Dept. of, Physics, Air University, Islamabad (Pakistan); Mirza, Sikander M. [Dept. of, Physics and Applied Mathematics, Pakistan Institute of Engineering and Applied Sciences, Post Office Nilore, Islamabad (Pakistan)

    2017-06-15

    This work looks at the effect of changes in kinetic parameters on simultaneous reactivity insertions and beam tube flooding in a typical material testing reactor-type research reactor with low enriched high density (U{sub 3}Si{sub 2}-Al) fuel. Using a modified PARET code, various ramp reactivity insertions (from $0.1/0.5 s to $1.3/0.5 s) plus beam tube flooding ($0.5/0.25 s) accidents under uncontrolled conditions were analyzed to find their effects on peak power, net reactivity, and temperature. Then, the effects of changes in kinetic parameters including the Doppler coefficient, prompt neutron lifetime, and delayed neutron fractions on simultaneous reactivity insertion and beam tube flooding accidents were analyzed. Results show that the power peak values are significantly sensitive to the Doppler coefficient of the system in coupled accidents. The material testing reactor-type system under such a coupled accident is not very sensitive to changes in the prompt neutron life time; the core under such a coupled transient is not very sensitive to changes in the effective delayed neutron fraction.

  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. Review of activities of Research Association of High Temperature Gas Cooled Reactor Plant (RAHP)

    Energy Technology Data Exchange (ETDEWEB)

    An, Shigehiro [Research Association of High Temperature Gas Cooled Reactor Plant RAHP, University of Tokyo, Tokyo (Japan); Tsuchie, Yasuo [Research Association of High Temperature Gas Cooled Reactor Plant RAHP, The Japan Atomic Power Co. JAPC, Tokyo (Japan)

    1998-09-01

    The Research Association of the HTGR Plant (RAHP) is the sole research association in the private or industrial sector of Japan with respect to HTGR Plants. It was established in 1985, composing of professors, representatives of electric power companies, and fabricators of nuclear plant and fuels Activities in these years were to analyze world trends of R and D, to identify techno-economical issues to be cleared, to set-up fundamental development strategies, and to put the results of the studies into actions towards commercialization of the HTGR. Conclusions obtained through the activities so far are: (1) From the view point of effective use of energy and reduction of environmental impacts on a global scale, development of nuclear power is essential, in particular of the HTGR, because of its very highly inherent safety and feasibility of high temperature heat uses. The role of the HTGR is inter-complementary with those of LWR and FBR; (2) Future subjects on the HTGR are technical demonstration of its unique characteristics, economic prospects, public acceptance (PA) and industrial acceptance, R and D through international cooperation and share in role, and successful realization of demonstration plant(s). RAHP is to start a survey on HTGR from nuclear fuel cycle point of view to have a better outlook on future needs of high temperature heat uses. 6 refs.

  10. Nuclear Fission Reactor Safety Research in FP7 and future perspectives

    CERN Document Server

    Garbil, Roger

    2014-01-01

    The European Union (ЕU) has defined in the Europe 2020 strategy and 2050 Energy Roadmap its long-term vision for establishing a secure, sustainable and competitive energy system and setting up legally binding targets by 2020 for reducing greenhouse emissions, by increasing energy efficiency and the share of renewable energy sources while including a significant share from nuclear fission. Nuclear energy can enable the further reduction in harmful emissions and can contribute to the EU’s competitive energy system, security of supply and independence from fossil fuels. Nuclear fission is a valuable option for those 14 EU countries that promote its use as part of their national energy mix. The European Group on Ethics in Science and New Technologies (EGE) adopted its Opinion No.27 ‘An ethical framework for assessing research, production and use of energy’ and proposed an integrated ethics approach for the research, production and use of energy in the EU, seeking equilibrium among four criteria – access ...

  11. DOE-NE Light Water Reactor Sustainability Program and EPRI Long-Term Operations Program. Joint Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Williams, Don

    2014-04-01

    Nuclear power has contributed almost 20% of the total amount of electricity generated in the United States over the past two decades. High capacity factors and low operating costs make nuclear power plants (NPPs) some of the most economical power generators available. Further, nuclear power remains the single largest contributor (nearly 70%) of non-greenhouse gas-emitting electric power generation in the United States. Even when major refurbishments are performed to extend operating life, these plants continue to represent cost-effective, low-carbon assets to the nation’s electrical generation capability. By the end of 2014, about one-third of the existing domestic fleet will have passed their 40th anniversary of power operations, and about one-half of the fleet will reach the same 40-year mark within this decade. Recognizing the challenges associated with pursuing extended service life of commercial nuclear power plants, the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE) and the Electric Power Research Institute (EPRI) have established separate but complementary research and development programs (DOE-NE’s Light Water Reactor Sustainability [LWRS] Program and EPRI’s Long-Term Operations [LTO] Program) to address these challenges. To ensure that a proper linkage is maintained between the programs, DOE-NE and EPRI executed a memorandum of understanding in late 2010 to “establish guiding principles under which research activities (between LWRS and LTO) could be coordinated to the benefit of both parties.” This document represents the third annual revision to the initial version (March 2011) of the plan as called for in the memorandum of understanding.

  12. New reactor type proposed

    CERN Multimedia

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  13. Conditioning of pretreated LLW generated by the decontamination of VVR-S nuclear research reactor primary circuit

    Energy Technology Data Exchange (ETDEWEB)

    Nicu, Mihaela I.; Ionascu, Laura A.; Dragolici, Felicia N.; Turcanu, Corneliu N.; Rotarescu, Gheorghe Gh.; Dogaru, Gheorghe C. [' Horia Hulubei' National Institute for Nuclear Physics and Engineering Magurele, Ilfov (Romania)

    2013-07-01

    Concentration of complexing acids solutions (oxalic acid, tartaric acid and citric acid) used in radioactive decontamination by chemical method affects both the mechanical stability and the chemical stability of cement matrix. The paper presents the works performed related to the chemical pretreatment of these organic acids solutions using as neutralizing agent Ca(OH){sub 2}. In this way it was possible to increase the concentration of organic acids solutions used and the soluble complex radionuclides passing in chemical precipitates, these affecting in a smaller manner the mechanical stability of the cement matrix. The chemical pretreatment the effluents improve the precipitation and conditioning performances by cementation. Were prepared compositions with complexing agents and compositions for oxidative degradation tests to simulate the concentrations of secondary radioactive waste obtained from the primary circuit decontamination of the VVR-S research reactor. It has been studied the influence of chemical pretreatment of complexing acids solutions of different concentrations on the setting time. Also it was determined the compressive strength of mortar samples in which were embedded these solutions of chemically pretreated organic acids. The results shown that an optimum cement - solution ratio doesn't have a significant impact on the setting time or on the mechanical properties. (authors)

  14. Measurement of neutron spectra in a silicon filtered neutron beam using stilbene detectors at the LVR-15 research reactor.

    Science.gov (United States)

    Košťál, Michal; Šoltés, Jaroslav; Viererbl, Ladislav; Matěj, Zdeněk; Cvachovec, František; Rypar, Vojtěch; Losa, Evžen

    2017-10-01

    A well-defined neutron spectrum is an essential tool for calibration and tests of spectrometry and dosimetry detectors, and evaluation methods for spectra processing. Many of the nowadays used neutron standards are calibrated against a fission spectrum which has a rather smooth energy dependence. In recent time, at the LVR-15 research reactor in Rez, an alternative approach was tested for the needs of fast neutron spectrometry detector calibration. This process comprises detector tests in a neutron beam, filtered by one meter of single-crystalline silicon, which contains several significant peaks in the fast neutron energy range. Tests in such neutron field can possibly reveal specific problems in the deconvolution matrix of the detection system, which may stay hidden in fields with a smooth structure and can provide a tool for a proper energy calibration. Test with several stilbene scintillator crystals in two different beam configurations supplemented by Monte-Carlo transport calculations have been carried out. The results have shown a high level of agreement between the experimental data and simulation, proving thus the accuracy of used deconvolution matrix. The chosen approach can, thus, provide a well-defined neutron reference field with a peaked structure for further tests of spectra evaluation methods and scintillation detector energy calibration. Copyright © 2017 Elsevier Ltd. All rights reserved.

  15. Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

    Energy Technology Data Exchange (ETDEWEB)

    Zamani, M. [National Radiation Protection Department - NRPD, Atomic Energy Organization of Iran - AEOI, Tehran (Iran, Islamic Republic of); End of North Kargar st, Atomic Energy Organization of Iran, P.O. Box: 14155-1339, Tehran (Iran, Islamic Republic of); Kasesaz, Y.; Khalafi, H.; Shayesteh, M. [Radiation Application School, Nuclear Science and Technology Research Institute, AEOI, Tehran (Iran, Islamic Republic of)

    2015-07-01

    In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials with different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)

  16. Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation

    Energy Technology Data Exchange (ETDEWEB)

    Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

    2008-07-01

    On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their high-enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.

  17. Detailed Design of the Safety Residual Heat Removal System and a Circulation Pump for the KIJANG Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Hyungi; Seo, Kyoungwoo; Kim, Seonghoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Primary cooling system (PCS) circulates the coolant from the reactor core to the heat exchanger. Therefore the heat generated from the fuel assembly in the reactor core is removed continuously. The PCS is designed based on the required thermal design flow rate of the reactor core, uncertainty of measuring instruments and the safe functions. Primary coolant is generally dumped into the pool and goes to the reactor core through the flow guide. The fission heat generated from the fuel assembly is transferred to the coolant, and then heated coolant goes to the PCS equipment room in order to remove the heat through the heat exchanger. SRHSR is designed based on the required flow rate and system constraints. Centrifugal pump of Case 1 with a non-dimensional specific speed of 0.97 [-] and specific diameter of 3.33 [-] is chosen as the SRHRS pump for the KJRR.

  18. Implementation of MPC and A systems at the VNIIEF research (reactor) site

    Energy Technology Data Exchange (ETDEWEB)

    Skripka, G.; Koloyartsev, V.; Mikijchuk, N. [Rossijskij Federal`nyj Yadernyj Tsentr, Arzamas (Russian Federation). Nauchno-Issledovatel`skij Inst. Ehksperimental`noj Fiziki] [and others

    1996-09-01

    As part of the US-Russian Lab-to-Lab program for strengthening nuclear material protection, control, and accounting (MPC and A), an integrated MPC and A system is being installed at a large site within the VNIIEF complex to upgrade safeguards on the nuclear material stored and used at this site. In addition to storage facilities, the site houses a number of critical facilities at which nuclear physics research is conducted. The design of the MPC and A system is based on the test bed work presented earlier and provides the functions of nondestructive measurements for plutonium and highly enriched uranium, item control, personnel access control, radiation portal monitoring, and computed on-line accounting. The system controls, monitors, and accounts for nuclear material and people as the material moves through three material balance areas, a measurement room, and a number of control points. It also assists with physical inventory taking. The instrumentation used to implement these functions will include US commercial equipment as well as Russian-designed and -fabricated items.

  19. CONVECTION REACTOR

    Science.gov (United States)

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  20. Accelerating research into bio-based FDCA-polyesters by using small scale parallel film reactors.

    Science.gov (United States)

    Gruter, Gert-Jan M; Sipos, Laszlo; Adrianus Dam, Matheus

    2012-02-01

    High Throughput experimentation has been well established as a tool in early stage catalyst development and catalyst and process scale-up today. One of the more challenging areas of catalytic research is polymer catalysis. The main difference with most non-polymer catalytic conversions is the fact that the product is not a well defined molecule and the catalytic performance cannot be easily expressed only in terms of catalyst activity and selectivity. In polymerization reactions, polymer chains are formed that can have various lengths (resulting in a molecular weight distribution rather than a defined molecular weight), that can have different compositions (when random or block co-polymers are produced), that can have cross-linking (often significantly affecting physical properties), that can have different endgroups (often affecting subsequent processing steps) and several other variations. In addition, for polyolefins, mass and heat transfer, oxygen and moisture sensitivity, stereoregularity and many other intrinsic features make relevant high throughput screening in this field an incredible challenge. For polycondensation reactions performed in the melt often the viscosity becomes already high at modest molecular weights, which greatly influences mass transfer of the condensation product (often water or methanol). When reactions become mass transfer limited, catalyst performance comparison is often no longer relevant. This however does not mean that relevant experiments for these application areas cannot be performed on small scale. Relevant catalyst screening experiments for polycondensation reactions can be performed in very efficient small scale parallel equipment. Both transesterification and polycondensation as well as post condensation through solid-stating in parallel equipment have been developed. Next to polymer synthesis, polymer characterization also needs to be accelerated without making concessions to quality in order to draw relevant conclusions.

  1. Conceptual analyses of neutronic and equilibrium refueling parameters to develop a cost-effective multi-purpose pool-type research reactor using WIMSD and CITVAP codes

    Energy Technology Data Exchange (ETDEWEB)

    Hedayat, Afshin, E-mail: ahedayat@aeoi.org.ir

    2016-12-01

    Highlights: • Introducing a high-beneficent and low-cost multipurpose research reactor. • High technical documents and standard safety issues are introduced coherently. • High effective conceptual neutronic analyses and fuel management strategy. • Gaining high score design criteria and safety margins via 3-D core modeling. • Capacity and capability to produce all medical and industrial radioisotopes. - Abstract: In this paper, neutronic and equilibrium refueling parameters of a multi-purpose cost-effective research reactor have been studied and analyzed. It has been tried to provide periodic and long-term requirements of the irradiating applications coherently. The WIMSD5B and CITVAP codes are used to calculate neutronic parameters and simulate fuel management strategy. The used nuclear data, codes, and calculating methods have been severally benchmarked and verified, successfully. Fundamental concepts, design criteria, and safety issues are introduced and discussed, coherently. Design criteria are selected to gain the most economic benefits per capital costs via minimum required reactor power. Accurate, fast and simplified models have been tried for an integrated decision making and analyses using deterministic codes. Core management, power effects, fuel consumption and burn up effects, and also a complete simulation of the fuel management strategy are presented and analyzed. Results show that the supposed reactor core design can be promisingly suitable in accordance with the commercial multi-purpose irradiating applications. It also retains Operating Limits and Conditions (OLCs) due to standard safety issues, conservatively where safety parameters are calculated using best estimate tools. Such reactor core configuration and integrated refueling task can effectively enhance the Quality Assurance (QA) of the general irradiating applications of the current MTR within their power limits and corresponding OLCs.

  2. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  3. FASTER Test Reactor Preconceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-31

    The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  4. DEVELOPMENT OF CONSTRUCTION OF A CATALYTIC REACTOR FOR METHANE OXIDISING IN VENTILATION AIR IN COAL MINES AND THE RESEARCH ON INTEGRATED “HEAT PIPE” RECUPERATOR

    Directory of Open Access Journals (Sweden)

    Lech Hys

    2013-04-01

    Full Text Available The article presents the analysis whose result is the selection of appropriate design and construction of a monolithic CMR reactor intended for oxidising methane from ventilation air in coal mines. The description of “heat-pipe” recuperator cooperating with the reactor was also presented. The research was mainly aimed at verifying the compliance with the work of autothermity premise obtained by the return of part of heat from catalytic reaction. The result of research was to define the range volumetric fume expense ensuring autothermity and the definition of maximum recuperator efficiency. The range of volumetric expense was 18–25 m3/h and maximum value of efficiency coefficient was η = 0.50 for the volumetric expense of 18 m3/h.

  5. Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stage 11 and 12. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Sunde, C.; Pazsit, I.; Demaziere, C.; Dahl, O.; Mileshina, L. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Nuclear Engineering

    2006-06-15

    This report describes the results obtained during Stage 11 and 12 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. The long-term goals are elaborated in more detail in e.g. the Final Reports of stage 1 and 2 (SKI Report 95:14 and 96:50). Higher order eigenfunctions and eigenvalues of the diffusion equation, describing a static core, have lost their significance when doing calculations in realistic systems, since there are usually made by nodal methods or other direct numerical techniques. However, there are situations when knowledge of the higher order modes is still useful. Such case is the separation of the global and regional flux oscillations in the case of BWR instability. Another case is the investigation of the stability properties of large inhomogeneous cores, which is usually quantified with the so-called eigenvalue separation, ES=1/k{sub 1} - 1/k{sub 0} where k{sub 0} and k{sub 1}, are the fundamental and first higher order eigenvalues, respectively. Numerical codes used for ICMF calculations usually do not have the option of calculating the higher order eigenvalues and eigenfunctions. In idealised systems, these can be calculated with semi-analytical methods. In order to get insight into the characteristics of the higher order eigenmodes of the neutron flux and the adjoint, these were calculated in a reflected homogeneous system with two-group theory. The adjoint eigenmodes are necessary if an orthogonality property needs to be used, such as the separation of the modes from a flux shape which is a sum of several eigenmodes. At the same time, the so-called numerical noise simulator, developed at the Dept., was extended such that it became suitable for the calculation of the higher order eigenmodes, both direct and adjoint ones. This simulator can treat real inhomogeneous cores, with an input deck compatible with that of SIMULATE. The simulator was used to calculate the

  6. A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

    Energy Technology Data Exchange (ETDEWEB)

    Kamal, I., E-mail: imtiaz-kamal26@yahoo.com; Yunus, S. M., E-mail: yunussm11@yahoo.com; Datta, T. K., E-mail: tk-datta4@yahoo.com; Zakaria, A. K. M.; Das, A. K.; Aktar, S.; Hossain, S. [Institute of Nuclear Science and Technology, Atomic Energy Research Establishment, Ganakbari, Savar, Dhaka (Bangladesh); Berliner, R., E-mail: RB@instrumentationAssociates.com [Instrumentation Associates, Durham, North Carolina (United States); Yelon, W. B., E-mail: yelonwb@hotmail.com [Oak Ridge National Laboratory, Oak Ridge, Tennessee (United States)

    2016-07-12

    A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with a large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6 mm thickness each. The monochromator design was optimized to provide maximum flux on 3 mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30° (2θ) at each step and covers 120° in 4 steps. When the detector is placed at 1.6 m it subtends 20° at each step and covers 120° in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art technology

  7. Optimization study for an epithermal neutron beam for boron neutron capture therapy at the University of Virginia Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Jr., Thomas Dean [Univ. of Virginia, Charlottesville, VA (United States)

    1995-05-01

    The non-surgical brain cancer treatment modality, Boron Neutron Capture Therapy (BNCT), requires the use of an epithermal neutron beam. This purpose of this thesis was to design an epithermal neutron beam at the University of Virginia Research Reactor (UVAR) suitable for BNCT applications. A suitable epithermal neutron beam for BNCT must have minimal fast neutron and gamma radiation contamination, and yet retain an appreciable intensity. The low power of the UVAR core makes reaching a balance between beam quality and intensity a very challenging design endeavor. The MCNP monte carlo neutron transport code was used to develop an equivalent core radiation source, and to perform the subsequent neutron transport calculations necessary for beam model analysis and development. The code accuracy was validated by benchmarking output against experimental criticality measurements. An epithermal beam was designed for the UVAR, with performance characteristics comparable to beams at facilities with cores of higher power. The epithermal neutron intensity of this beam is 2.2 x 108 n/cm2 • s. The fast neutron and gamma radiation KERMA factors are 10 x 10-11cGy•cm2/nepi and 20 x 10-11 cGy•cm2/nepi , respectively, and the current-to-flux ratio is 0.85. This thesis has shown that the UVAR has the capability to provide BNCT treatments, however the performance characteristics of the final beam of this study were limited by the low core power.

  8. A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

    Science.gov (United States)</