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Sample records for anuclear research reactor

  1. Research reactors

    SciTech Connect

    Tonneson, L.C.; Fox, G.J.

    1996-04-01

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world`s research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted.

  2. Research reactors - an overview

    SciTech Connect

    West, C.D.

    1997-03-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

  3. Research Reactor Benchmarks

    SciTech Connect

    Ravnik, Matjaz; Jeraj, Robert

    2003-09-15

    A criticality benchmark experiment performed at the Jozef Stefan Institute TRIGA Mark II research reactor is described. This experiment and its evaluation are given as examples of benchmark experiments at research reactors. For this reason the differences and possible problems compared to other benchmark experiments are particularly emphasized. General guidelines for performing criticality benchmarks in research reactors are given. The criticality benchmark experiment was performed in a normal operating reactor core using commercially available fresh 20% enriched fuel elements containing 12 wt% uranium in uranium-zirconium hydride fuel material. Experimental conditions to minimize experimental errors and to enhance computer modeling accuracy are described. Uncertainties in multiplication factor due to fuel composition and geometry data are analyzed by sensitivity analysis. The simplifications in the benchmark model compared to the actual geometry are evaluated. Sample benchmark calculations with the MCNP and KENO Monte Carlo codes are given.

  4. Reactor Safety Research Programs

    SciTech Connect

    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.

  5. Reactor Safety Research Programs

    SciTech Connect

    Dotson, CW

    1980-08-01

    This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty 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 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 the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  6. Gaseous fuel reactor research

    NASA Technical Reports Server (NTRS)

    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.

  7. Research reactor fork users manual

    SciTech Connect

    Hsue, S.T.; Menlove, H.O.; Bosler, G.E.; Dye, H.R.; Walton, G.; Halbig, J.K.; Siebelist, R.

    1993-11-01

    This manual describes the design features and operating characteristics of the research reactor fork. The system includes an ion chamber for gross gamma-ray counting, fission chambers for neutron counting, and a collimated high-resolution spectroscopy system for gamma-ray measurements. The neutron and ion chamber measurements are designed to be made underwater in spent-fuel cooling ponds. The neutron and gamma-ray detectors have been designed with high efficiencies to accommodate the relatively low emission rates of neutrons and gamma rays from low-burnup, research-type reactor fuel. This manual presents the design, performance, and test results for the system.

  8. International Research Reactor Decommissioning Project

    SciTech Connect

    Leopando, Leonardo; Warnecke, Ernst

    2008-01-15

    Many research reactors have been or will be shut down and are candidates for decommissioning. Most of the respective countries neither have a decommissioning policy nor the required expertise and funds to effectively implement a decommissioning project. The IAEA established the Research Reactor Decommissioning Demonstration Project (R{sup 2}D{sup 2}P) to help answer this need. It was agreed to involve the Philippine Research Reactor (PRR-1) as model reactor to demonstrate 'hands-on' experience as it is just starting the decommissioning process. Other facilities may be included in the project as they fit into the scope of R{sup 2}D{sup 2}P and complement to the PRR-1 decommissioning activities. The key outcome of the R{sup 2}D{sup 2}P will be the decommissioning of the PRR-1 reactor. On the way to this final goal the preparation of safety related documents (i.e., decommissioning plan, environmental impact assessment, safety analysis report, health and safety plan, cost estimate, etc.) and the licensing process as well as the actual dismantling activities could provide a model to other countries involved in the project. It is expected that the R{sup 2}D{sup 2}P would initiate activities related to planning and funding of decommissioning activities in the participating countries if that has not yet been done.

  9. Research Program of a Super Fast Reactor

    SciTech Connect

    Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki; Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki; GOTO, Shoji

    2006-07-01

    Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is not breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)

  10. Reactivity Transients in Nuclear Research Reactors

    Energy Science and Technology Software Center (ESTSC)

    2015-01-01

    Version 01 AIREMOD-RR is a point kinetics code which can simulate fast transients in nuclear research reactor cores. It can also be used for theoretical reactor dynamics studies. It is used for research reactor kinetic analysis and provides a point neutron kinetic capability. The thermal hydraulic behavior is governed by a one-dimensional heat balance equation. The calculations are restricted to a single equivalent unit cell which consists of fuel, clad and coolant.

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

    SciTech Connect

    Douglas Morrell

    2011-03-01

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

  12. Utilisation of British University Research Reactors.

    ERIC Educational Resources Information Center

    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…

  13. Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

    SciTech Connect

    Hauptman, H.M.; Petro, J.N.; Jacobi, O.

    1995-04-01

    This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed.

  14. Gaseous fuel nuclear reactor research

    NASA Technical Reports Server (NTRS)

    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.

  15. Supply of enriched uranium for research reactors

    SciTech Connect

    Mueller, H.

    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.

  16. Fuel elements of research reactor CM

    SciTech Connect

    Kozlov, A.V.; Morozov, A.V.; Vatulin, A.V.; Ershov, S.A.

    2013-07-01

    In 1961 the CM research reactor was commissioned at the Research Institute of Atomic Reactors (Dimitrovgrad, Russia), it was intended to carry on investigations and the production of transuranium nuclides. The reactor is of a tank type. Original fuel assembly contained plate fuels that were spaced with vanes and corrugated bands. Nickel was used as a cladding material, fuel meat was produced from UO{sub 2} + electrolytic nickel composition. Fuel plates have been replaced by self-spacing cross-shaped dispersion fuels clad in stainless steel. In 2005 the reactor was updated. The purpose of this updating was to increase the quantity of irradiation channels in the reactor core and to improve the neutron balance. The updating was implemented at the expense of 20 % reduction in the quantity of fuel elements in the core which released a space for extra channels and decreased the mass of structural materials in the core. The updated reactor is loaded with modified standard fuel elements with 20 % higher uranium masses. At the same time stainless steel in fuel assembly shrouds was substituted by zirconium alloy. Today in progress are investigations and work to promote the second stage of reactor updating that involve developments of cross-shaped fuel elements having low neutron absorption matrix materials. This article gives an historical account of the design and main technical changes that occurred for the CM reactor since its commissioning.

  17. Research on plasma core reactors

    NASA Technical Reports Server (NTRS)

    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.

  18. Probabilistic Safety Assessment of Tehran Research Reactor

    SciTech Connect

    Hosseini, Seyed Mohammad Hadi; Nematollahi, Mohammad Reza; Sepanloo, Kamran

    2004-07-01

    Probabilistic Safety Assessment (PSA) application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. In this paper the application of the Probabilistic Safety Assessment to the Tehran Research Reactor (TRR) is presented. The level 1 PSA application involved: Familiarization with the plant, selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development and dependent failure analysis. Each of the steps of the analysis given above is discussed with highlights from the selected results. Quantification of the constructed models is done using SAPHIRE software. This Study shows that the obtained core damage frequency for Tehran Research Reactor (8.368 E-6 per year) well meets the IAEA criterion for existing nuclear power plants (1E-4). But safety improvement suggestions are offered to decrease the most probable accidents. (authors)

  19. Corrosion Minimization for Research Reactor Fuel

    SciTech Connect

    Eric Shaber; Gerard Hofman

    2005-06-01

    Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

  20. Facility modernization Annular Core Research Reactor

    SciTech Connect

    Morris, F.M.; Luera, T.F.; McCrory, F.M.; Nelson, D.A.; Trowbridge, F.R.; Wold, S.A.

    1990-07-01

    The Annular Core Research Reactor (ACRR) has undergone numerous modifications since its conception in response to program needs. The original reactor fuel, which was special U-ZrH TRIGA fuel designed primarily for pulsing, has been replaced with a higher pulsing capacity BeO fuel. Other advanced operating modes which use this increased capability, in addition to the pulse and steady state, have been incorporated to tailor power histories and fluences to the experiments. Various experimental facilities have been developed that range from a radiography facility to a 50 cm diameter External Fuel Ring Cavity (FREC) using 180 of the original ZrH fuel elements. Currently a digital reactor console is being produced with GA, which will give enhanced monitoring capabilities of the reactor parameters while leaving the safety-related shutdown functions with analog technology. (author)

  1. Fuel behavior comparison for a research reactor

    NASA Astrophysics Data System (ADS)

    Negut, Gh.; Mladin, M.; Prisecaru, I.; Danila, N.

    2006-06-01

    The paper presents the behavior and properties analysis of the low enriched uranium fuel, which will be loaded in the Romanian TRIGA 14 MW steady state research reactor compared with the original high enriched uranium fuel. The high and low enriched uranium fuels have similar thermal properties, but different nuclear properties. The research reactor core was modeled with both fuel materials and the reactor behavior was studied during a reactivity insertion accident. The thermal hydraulic analysis results are compared with that obtained from the safety analysis report for high enriched uranium fuel core. The low enriched uranium fuel shows a good behavior during reactivity insertion accident and a revised safety analysis report will be made for the low enriched uranium fuel core.

  2. Trends in fusion reactor safety research

    NASA Astrophysics Data System (ADS)

    Herring, J. S.; Holland, D. F.; Piet, S. J.

    Fusion has the potential to be an attractive energy source. From the safety and environmental perspective, fusion must avoid concerns about catastrophic accidents and unsolvable waste disposal. In addition, fusion must achieve an acceptable level of risk from operational accidents that result in public exposure and economic loss. Finally, fusion reactors must control routine radioactive effluent, particularly tritium. Major progress in achieving this potential rests on development of low-activation materials or alternative fuels. The safety and performance of various material choices and fuels for commercial fusion reactors can be investigated relatively inexpensively through reactor design studies. These studies bring together experts in a wide range of backgrounds and force the group to either agree on a reactor design or identify areas for further study. Fusion reactors will be complex, with distributed radioactive inventories. The next generation of experiments will be critical in demonstrating that acceptable levels of safe operation can be achieved. These machines will use materials which are available today and for which a large database exists (e.g., for 316 stainless steel). Researchers have developed a good understanding of the risks associated with operation of these devices. Specifically, consequences from coolant system failures, loss of vacuum events, tritium releases, and liquid metal reactions have been studied. Recent studies go beyond next step designs and investigate commercial reactor concerns including tritium release and liquid metal reactions.

  3. Austrian contributions to reactor safety research

    NASA Astrophysics Data System (ADS)

    Sdouz, Gert

    1992-03-01

    An overview of Austrian contributions to reactor safety research is given. Starting with licensing for the power plant Zwenkendorf and participation in the safety projects PBF (Power Burst Facility) and LOFT (Loss Of Fluid Test), the work shifted later to participation in the study of international standard problems of the OECD and the IAEA. Since 1988 the center of activity is the calculation of the source term behavior of accident sequences, especially for reactors of Russian types. As an illustration the results of a source term calculation for a TMLB' sequence are presented.

  4. Radionuclide release from research reactor spent fuel

    NASA Astrophysics Data System (ADS)

    Curtius, H.; Kaiser, G.; Müller, E.; Bosbach, D.

    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 2 fuel matrix. Special fuel-types (mostly dispersed fuels, high enriched in 235U, cladded with aluminium) are used in German research reactors, whereas in German nuclear power plants, UO 2-fuel (LWR fuel, enrichment in 235U 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 Jülich 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 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 x-Al and U 3Si 2-Al) was studied in 400 mL MgCl 2-rich salt brine in the presence of Fe 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) 3(s) and Eu(OH) 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(IV) oxyhydroxides species due to radiolytic effects cannot

  5. 78 FR 58575 - Review of Experiments for Research Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    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...

  6. Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program

    SciTech Connect

    Not Available

    1987-05-01

    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 accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance 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 technology base supporting licensing decisions.

  7. Monte Carlo modelling of TRIGA research reactor

    NASA Astrophysics Data System (ADS)

    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.

  8. TRIGA research reactor activities around the world

    SciTech Connect

    Chesworth, R.H.; Razvi, J.; Whittemore, W.L. )

    1991-11-01

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

  9. Research reactor de-fueling and fuel shipment

    SciTech Connect

    Ice, R.D.; Jawdeh, E.; Strydom, J.

    1998-08-01

    Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures.

  10. Global estimation of potential unreported plutonium in thermal research reactors

    SciTech Connect

    Dreicer, J.S.; Rutherford, D.A.

    1996-09-01

    As of November, 1993, 303 research reactors (research, test, training, prototype, and electricity producing) were operational worldwide; 155 of these were in non-nuclear weapon states. Of these 155 research reactors, 80 are thermal reactors that have a power rating of 1 MW(th) or greater and could be utilized to produce plutonium. A previously published study on the unreported plutonium production of six research reactors indicates that a minimum reactor power of 40 MW (th) is required to make a significant quantity (SQ), 8 kg, of fissile plutonium per year by unreported irradiations. As part of the Global Nuclear Material Control Model effort, we determined an upper bound on the maximum possible quantity of plutonium that could be produced by the 80 thermal research reactors in the non-nuclear weapon states (NNWS). We estimate that in one year a maximum of roughly one quarter of a metric ton (250 kg) of plutonium could be produced in these 80 NNWS thermal research reactors based on their reported power output. We have calculated the quantity of plutonium and the number of years that would be required to produce an SQ of plutonium in the 80 thermal research reactors and aggregated by NNWS. A safeguards approach for multiple thermal research reactors that can produce less than 1 SQ per year should be conducted in association with further developing a safeguards and design information reverification approach for states that have multiple research reactors.

  11. Health physics research reactor reference dosimetry

    SciTech Connect

    Sims, C.S.; Ragan, G.E.

    1987-06-01

    Reference neutron dosimetry is developed for the Health Physics Research Reactor (HPRR) in the new operational configuration directly above its storage pit. This operational change was physically made early in CY 1985. The new reference dosimetry considered in this document is referred to as the 1986 HPRR reference dosimetry and it replaces any and all HPRR reference documents or papers issued prior to 1986. Reference dosimetry is developed for the unshielded HPRR as well as for the reactor with each of five different shield types and configurations. The reference dosimetry is presented in terms of three different dose and six different dose equivalent reporting conventions. These reporting conventions cover most of those in current use by dosimetrists worldwide. In addition to the reference neutron dosimetry, this document contains other useful dosimetry-related data for the HPRR in its new configuration. These data include dose-distance measurements and calculations, gamma dose measurements, neutron-to-gamma ratios, ''9-to-3 inch'' ratios, threshold detector unit measurements, 56-group neutron energy spectra, sulfur fluence measurements, and details concerning HPRR shields. 26 refs., 11 figs., 31 tabs.

  12. Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

    SciTech Connect

    Huy, N.Q.; Thong, H.V.; Khang, N.P.

    1994-12-31

    The Dalat nuclear research reactor was reconstructed from the TRIGA Mark II reactor, built in 1963 with a nominal power of 250 kW, and reached its planned nominal power of 500 kW for the first time in February 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at a deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc.

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

    SciTech Connect

    Rosenthal, Murray Wilford

    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.

  14. Diversion assumptions for high-powered research reactors

    SciTech Connect

    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.

  15. Disassembly of the Research Reactor FRJ-1 (MERLIN)

    SciTech Connect

    Stahn, B.; Poeppinghaus, J.; Cremer, J.

    2002-02-25

    This report describes the past steps of dismantling the research reactor FRJ-1 (MERLIN) and, moreover, provides an outlook on future dismantling with the ultimate aim of a ''green field site''. MERLIN is an abbreviation for MEDIUM ENERGY RESEARCH LIGHT WATER MODERATED INDUSTRIAL NUCLEAR REACTOR.

  16. REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

    SciTech Connect

    Nichols, T.; Beals, D.; Sternat, M.

    2011-07-18

    Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Many research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical

  17. REACTOR PHYSICS MODELING OF SPENT NUCLEAR RESEARCH REACTOR FUEL FOR SNM ATTRIBUTION AND NUCLEAR FORENSICS

    SciTech Connect

    Sternat, M.; Beals, D.; Webb, R.; Nichols, T.

    2010-06-09

    Nuclear research reactors are the least safeguarded type of reactor; in some cases this may be attributed to low risk and in most cases it is due to difficulty from dynamic operation. Research reactors vary greatly in size, fuel type, enrichment, power and burnup providing a significant challenge to any standardized safeguard system. If a whole fuel assembly was interdicted, based on geometry and other traditional forensics work, one could identify the material's origin fairly accurately. If the material has been dispersed or reprocessed, in-depth reactor physics models may be used to help with the identification. Should there be a need to attribute research reactor fuel material, the Savannah River National Laboratory would perform radiochemical analysis of samples of the material as well as other non-destructive measurements. In depth reactor physics modeling would then be performed to compare to these measured results in an attempt to associate the measured results with various reactor parameters. Several reactor physics codes are being used and considered for this purpose, including: MONTEBURNS/ORIGEN/MCNP5, CINDER/MCNPX and WIMS. In attempt to identify reactor characteristics, such as time since shutdown, burnup, or power, various isotopes are used. Complexities arise when the inherent assumptions embedded in different reactor physics codes handle the isotopes differently and may quantify them to different levels of accuracy. A technical approach to modeling spent research reactor fuel begins at the assembly level upon acquiring detailed information of the reactor to be modeled. A single assembly is run using periodic boundary conditions to simulate an infinite lattice which may be repeatedly burned to produce input fuel isotopic vectors of various burnups for a core level model. A core level model will then be constructed using the assembly level results as inputs for the specific fuel shuffling pattern in an attempt to establish an equilibrium cycle. The

  18. Korea Research Reactor -1 & 2 Decommissioning Project in Korea

    SciTech Connect

    Park, S. K.; Chung, U. S.; Jung, K. J.; Park, J. H.

    2003-02-24

    Korea Research Reactor 1 (KRR-1), the first research reactor in Korea, has been operated since 1962, and the second one, Korea Research Reactor 2 (KRR-2) since 1972. The operation of both of them was phased out in 1995 due to their lifetime and operation of the new and more powerful research reactor, HANARO (High-flux Advanced Neutron Application Reactor; 30MW). Both are TRIGA Pool type reactors in which the cores are small self-contained units sitting in tanks filled with cooling water. The KRR-1 is a TRIGA Mark II, which could operate at a level of up to 250 kW. The second one, the KRR-2 is a TRIGA Mark III, which could operate at a level of up 2,000 kW. The decontamination and decommissioning (D & D) project of these two research reactors, the first D & D project in Korea, was started in January 1997 and will be completed to stage 3 by 2008. The aim of this decommissioning program is to decommission the KRR-1 & 2 reactors and to decontaminate the residual building structure s and the site to release them as unrestricted areas. KAERI (Korea Atomic Energy Research Institute) submitted the decommissioning plan and the environmental impact assessment reports to the Ministry of Science and Technology (MOST) for the license in December 1998, and was approved in November 2000.

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

    SciTech Connect

    Ludewig, H.; Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.; Lambert, J.; Hayes, S.; Sackett, J.; 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. A Potential NASA Research Reactor to Support NTR Development

    NASA Technical Reports Server (NTRS)

    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.

  1. Conversion of research and test reactors : status and current plans.

    SciTech Connect

    Roglans, J.; Staples, P.; Butler, N.; Nuclear Engineering Division

    2007-01-01

    The Office of Global Threat Reduction's (GTRI) Conversion Program develops technology necessary to enable the conversion of civilian facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets. The Conversion program mission supports the minimization and, to the extent possible, elimination of the use of HEU in civil nuclear applications by working to convert research reactors and radioisotope production processes to the use of LEU fuel and targets throughout the world. During the Program's 27 years of existence, 46 research reactors have been converted from HEU to LEU fuels and processes have been developed for producing the medical isotope Mo-99 with LEU targets. Under GTRI the Conversion Program has accelerated the schedules and plans for conversion of additional research reactors operating with HEU. Also the Program emphasizes the development of advanced high-density LEU fuels to enable further conversions. The Conversion program coordinates with the other program functions of GTRI, most notably the Removal function, which removes fresh and spent HEU fuel from countries around the world. This paper summarizes the current status and plans for conversion of research reactors, in the U.S. and abroad, the supporting fuel development activities, and the development of processes for medical isotope production with LEU targets. Nuclear research and test reactors worldwide have been in operation for over 60 years, supporting nuclear science and technology development, as well as providing an important role as a research tool in scientific fields including medicine, agriculture, industry, and basic research. Over 270 research reactors are currently operating in more than 50 countries. Starting in 1954, many research reactors outside the United States were provided under the Atoms for Peace initiative. Initial research reactors were fueled with low-enriched uranium (LEU) with a content of U235 of less than 20%. More advanced research

  2. JPL in-house fluidized-bed reactor research

    NASA Technical Reports Server (NTRS)

    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.

  3. The effective management of medical isotope production in research reactors

    SciTech Connect

    Drummond, D.T. )

    1993-01-01

    During the 50-yr history of the use of radioisotopes for medical applications, research reactors have played a pivotal role in the production of many if not most of the key products. The marriage between research reactors and production operations is subject to significant challenges on two fronts. The medical applications of the radioisotope products impose some unique constraints and requirements on the production process. In addition, the mandates and priorities of a research reactor are not always congruent with the demands of a production environment. This paper briefly reviews the historical development of medical isotope production, identifies the unique challenges facing this endeavor, and discusses the management of the relationship between the isotope producer and the research reactor operator. Finally, the key elements of a successful relationship are identified.

  4. DISMANTLING OF THE REACTOR BLOCK OF THE FRJ-1 RESEARCH REACTOR (MERLIN)

    SciTech Connect

    Stahn, B.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

    2003-02-27

    This report describes the past procedure in dismantling the reactor block of the FRJ-1 research reactor (MERLIN). Furthermore, it gives an outlook on future activities up to the final removal of the reactor block. MERLIN is an abbreviation for Medium Energy Research Light Water Moderated Industrial Nuclear Reactor. The FRJ-1 (MERLIN) was shut down in 1985 and the fuel elements removed from the facility. After dismantling the coolant loops and removing the reactor tank internals with subsequent draining of the reactor tank water, the first activities for dismantling the reactor block were carried out in summer 2001. The relevant license was granted in late July 2001 by the licensing authority specifying 8 incidental provisions. After dismantling the reactor extension (gates of the thermal columns and steel platforms surrounding the reactor block), a heavy-load platform including a casing around the reactor block was constructed. Two ventilation systems with a volume flow of 10,000 and 2 ,000 m3/h will, moreover, serve to avoid a spread of contamination. The reactor block will be dismantled in three phases divided according to upper, central and bottom sections. Dismantling the upper section started in August 2002. This section as well as the bottom section can probably be completely measured for clearance. For this reason, the activities have so far been carried out manually using mechanical and thermal techniques. The central section will probably have to be largely disposed of as radioactive waste. This is the region of the former reactor core in which the experimental devices are also integrated. Most of this work will probably have to be carried out by remote handling. More than 80 % of the dismantled materials of the reactor block can probably be measured for clearance. For this purpose, a clearance measurement device was taken into operation in the FRJ-1. On this occasion, the limits of clearance measurement have become evident. For concrete, which constitutes

  5. Background radiation measurements at high power research reactors

    NASA Astrophysics Data System (ADS)

    Ashenfelter, J.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Davee, D.; Dean, D.; Deichert, G.; Dolinski, M. J.; Dolph, J.; Dwyer, D. A.; Fan, S.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Green, M.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Jaffe, D. E.; Kettell, S.; Langford, T. J.; Littlejohn, B. R.; Martinez, D.; McKeown, R. D.; Morrell, S.; Mueller, P. E.; Mumm, H. P.; Napolitano, J.; Norcini, D.; Pushin, D.; Romero, E.; Rosero, R.; Saldana, L.; Seilhan, B. S.; Sharma, R.; Stemen, N. T.; Surukuchi, P. T.; Thompson, S. J.; Varner, R. L.; Wang, W.; Watson, S. M.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yeh, M.; Yen, Y.-R.; Zhang, C.; Zhang, X.

    2016-01-01

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  6. Background radiation measurements at high power research reactors

    DOE PAGESBeta

    Ashenfelter, J.; Yeh, M.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; et al

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the backgroundmore » fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.« less

  7. Background radiation measurements at high power research reactors

    SciTech Connect

    Ashenfelter, J.; Yeh, M.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Davee, D.; Dean, D.; Deichert, G.; Dolinski, M. J.; Dolph, J.; Dwyer, D. A.; Fan, S.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Green, M.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Jaffe, D. E.; Kettell, S.; Langford, T. J.; Littlejohn, B. R.; Martinez, D.; McKeown, R. D.; Morrell, S.; Mueller, P. E.; Mumm, H. P.; Napolitano, J.; Norcini, D.; Pushin, D.; Romero, E.; Rosero, R.; Saldana, L.; Seilhan, B. S.; Sharma, R.; Stemen, N. T.; Surukuchi, P. T.; Thompson, S. J.; Varner, R. L.; Wang, W.; Watson, S. M.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yen, Y. -R.; Zhang, C.; Zhang, X.

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  8. LEU conversion status of US research reactors, September 1996

    SciTech Connect

    Matos, J.E.

    1996-10-07

    This paper summarizes the conversion status of research and test reactors in the United States from the use of fuels containing highly- enriched uranium (HEU, greater than or equal to 20%) to the use of fuels containing low-enriched uranium (LEU, < 20%). Estimates of the uranium densities required for conversion are made for reactors with power levels greater than or equal to 1 MW that are not currently involved in the LEU conversion process.

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

    SciTech Connect

    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. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect

    Not Available

    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.

  11. Reactor pressure vessel structural integrity research

    SciTech Connect

    Pennell, W.E.; Corwin, W.R.

    1995-04-01

    Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallows surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT{sub NDT}) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on a shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties.

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

    SciTech Connect

    Not Available

    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.

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

    SciTech Connect

    Diamond, D. J.; Baek, J. S.; Hanson, A. L.; Cheng, L-Y; Brown, N.; Cuadra, A.

    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.

  14. Reduced-Enrichment Research and Test Reactor Program: Environmental assessment

    SciTech Connect

    Not Available

    1980-05-01

    The principal program objective and principal part of the proposed action is to improve the proliferation resistance of nuclear fuels used in research and test reactors by providing the technical means (through technical development, design, and testing) for reducing the uranium enrichment requirements of these fuels to substantially less than the 90 to 93% enrichment currently used. Operator acceptance of the reduced-enrichment-uranium (REU) fuel alternative will require minimizing of reactor performance reduction, fuel cycle cost increases, the number of new safety and licensing issues raised, and reactor and facility modifications. The other part of the proposed action is to assure the capability for commercial production and supply of REU fuel for use both in the US and abroad. The RERTR Program scope is limited to generic design studies, technical support to reactor operating organizations in preparing for conversions to REU fuels, fuel development, fuel demonstrations, and technical support for commercialization of REU fuels. This environmental assessment addresses the environmental consequences of RERTR Program activities and of specific conversions of typical reactors (the Ford Nuclear Reactor and one or two other to-be-designated demonstrations) to REU-fuel cycles, including domestic and international shipments of enriched uranium pertinent to the conduct of RERTR Program activities.

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

    SciTech Connect

    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.

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

    SciTech Connect

    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 January - March 1980

    SciTech Connect

    Hagen, C. M

    1980-10-01

    This document summarizes the work performed by Pacific Northwest Laboratory from January 1 through March 31, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty 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 steam generator tubes where serviceinduced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  18. Locating tritium sources in a research reactor building.

    PubMed

    Fukui, Masami

    2005-10-01

    Despite renovation of the D2O facility, tritium concentrations in the condensates of reactor room air showed tens of Bq mL before venting resumption on July 1997. This suggested the presence of tritium sources in the research reactor-containment building. An investigation was therefore initiated to locate the source and determine the distribution of tritium in the containment building. Air monitoring in the working area using a dish of water placed in the building suggested that the source of tritium was near the reactor core. Monitoring exhaust air from the two facilities (a cold neutron source and a D(2)O tank) showed high specific activity on the order of 10 Bq mL(-1), suggesting the presence of tritium in condensates near the reactor core. The major concern was whether the leakage of liquid deuterium (4 L) and heavy water (2 x 10(3) L) used as a moderator had occurred. The concentration of tritium in condensates has not increased over the past few years in either the exhaust line or working area, and the deuterium itself has not been found in the surrounding environment. The concentration of tritium measured using an ionization chamber after Ar decay was dependent on the thermal output of the research reactor, indicating that the tritium was produced by the irradiation process within shielding/moderator materials or cover gas with neutrons. PMID:16155451

  19. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect

    Not Available

    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.

  20. Iaea Activities Supporting the Applications of Research Reactors in 2013

    NASA Astrophysics Data System (ADS)

    Peld, Nathan D.; Ridikas, Danas

    2014-02-01

    As the underutilization of research reactors around the world persists as a primary topic of concern among facility owners and operators, the IAEA responded in 2013 with a broad range of activities to address the planning, execution and improvement of many experimental techniques. The revision of two critical documents for planning and diversifying a facility's portfolio of applications, TECDOC 1234 “The Applications of Research Reactors” and TECDOC 1212 “Strategic Planning for Research Reactors”, is in progress in order to keep this information relevant, corresponding to the dynamism of experimental techniques and research capabilities. Related to the latter TECDOC, the IAEA convened a meeting in 2013 for the expert review of a number of strategic plans submitted by research reactor operators in developing countries. A number of activities focusing on specific applications are either continuing or beginning as well. In neutron activation analysis, a joint round of inter-comparison proficiency testing sponsored by the IAEA Technical Cooperation Department will be completed, and facility progress in measurement accuracy is described. Also, a training workshop in neutron imaging and Coordinated Research Projects in reactor benchmarks, automation of neutron activation analysis and neutron beam techniques for material testing intend to advance these activities as more beneficial services to researchers and other users.

  1. RADIATION DOSIMETRY AT THE BNL HIGH FLUX BEAM REACTOR AND MEDICAL RESEARCH REACTOR.

    SciTech Connect

    HOLDEN,N.E.

    1999-09-10

    RADIATION DOSIMETRY MEASUREMENTS HAVE BEEN PERFORMED OVER A PERIOD OF MANY YEARS AT THE HIGH FLUX BEAM REACTOR (HFBR) AND THE MEDICAL RESEARCH REACTOR (BMRR) AT BROOKHAVEN NATIONAL LABORATORY TO PROVIDE INFORMATION ON THE ENERGY DISTRIBUTION OF THE NEUTRON FLUX, NEUTRON DOSE RATES, GAMMA-RAY FLUXES AND GAMMA-RAY DOSE RATES. THE MCNP PARTICLE TRANSPORT CODE PROVIDED MONTE CARLO RESULTS TO COMPARE WITH VARIOUS DOSIMETRY MEASUREMENTS PERFORMED AT THE EXPERIMENTAL PORTS, AT THE TREATMENT ROOMS AND IN THE THIMBLES AT BOTH HFBR AND BMRR.

  2. Technological Transfer from Research Nuclear Reactors to New Generation Nuclear Power Reactors

    SciTech Connect

    Radulescu, Laura; Pavelescu, Margarit

    2010-01-21

    The goal of this paper is the analysis of the technological transfer role in the nuclear field, with particular emphasis on nuclear reactors domain. The presentation is sustained by historical arguments. In this frame, it is very important to start with the achievements of the first nuclear systems, for instant those with natural uranium as fuel and heavy water as moderator, following in time through the history until the New Generation Nuclear Power Reactors.Starting with 1940, the accelerated development of the industry has implied the increase of the global demand for energy. In this respect, the nuclear energy could play an important role, being essentially an unlimited source of energy. However, the nuclear option faces the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide, a significant amount of experience has been accumulated during development, licensing, construction, and operation of nuclear power reactors. The experience gained is a strong basis for further improvements. Actually, the nuclear programs of many countries are addressing the development of advanced reactors, which are intended to have better economics, higher reliability, improved safety, and proliferation-resistant characteristics in order to overcome the current concerns about nuclear power. Advanced reactors, now under development, may help to meet the demand for energy power of both developed and developing countries as well as for district heating, desalination and for process heat.The paper gives historical examples that illustrate the steps pursued from first research nuclear reactors to present advanced power reactors. Emphasis was laid upon the fact that the progress is due to the great discoveries of the nuclear scientists using the technological transfer.

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

    SciTech Connect

    Heeger, Karsten M.

    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.

  4. MCNP/MCNPX model of the annular core research reactor.

    SciTech Connect

    DePriest, Kendall Russell; Cooper, Philip J.; Parma, Edward J., Jr.

    2006-10-01

    Many experimenters at the Annular Core Research Reactor (ACRR) have a need to predict the neutron/gamma environment prior to testing. In some cases, the neutron/gamma environment is needed to understand the test results after the completion of an experiment. In an effort to satisfy the needs of experimenters, a model of the ACRR was developed for use with the Monte Carlo N-Particle transport codes MCNP [Br03] and MCNPX [Wa02]. The model contains adjustable safety, transient, and control rods, several of the available spectrum-modifying cavity inserts, and placeholders for experiment packages. The ACRR model was constructed such that experiment package models can be easily placed in the reactor after being developed as stand-alone units. An addition to the 'standard' model allows the FREC-II cavity to be included in the calculations. This report presents the MCNP/MCNPX model of the ACRR. Comparisons are made between the model and the reactor for various configurations. Reactivity worth curves for the various reactor configurations are presented. Examples of reactivity worth calculations for a few experiment packages are presented along with the measured reactivity worth from the reactor test of the experiment packages. Finally, calculated neutron/gamma spectra are presented.

  5. Convective cooling in a pool-type research reactor

    NASA Astrophysics Data System (ADS)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

  6. A Research Reactor Concept to Support NTP Development

    NASA Technical Reports Server (NTRS)

    Eades, Michael J.; Blue, T. E.; Gerrish, Harold P.; Hardin, Leroy A.

    2014-01-01

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

  7. Fuels for research and test reactors, status review: July 1982

    SciTech Connect

    Stahl, D.

    1982-12-01

    A thorough review is provided on nuclear fuels for steady-state thermal research and test reactors. The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current US policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on US activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy, UZrH/sub x/, and UO/sub 2/ rod fuels. Among new fuels, those given major emphasis include H/sub 3/Si-Al dispersion and UO/sub 2/ caramel plate fuels.

  8. A probabilistic safety analysis of incidents in nuclear research reactors.

    PubMed

    Lopes, Valdir Maciel; Agostinho Angelo Sordi, Gian Maria; Moralles, Mauricio; Filho, Tufic Madi

    2012-06-01

    This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64. PMID:22021060

  9. Core conversion anaylses for the Portuguese Research Reactor.

    SciTech Connect

    Matos, J. E.; Stevens, J. G.; Feldman, E. E.; Stillman, J. A.; Dunn, F. E.; Kalimullah, K.; Marques, J. G.; Barradas, N. P.; Ramos, A .R.; Kling, A.; Inst. Tecnologico e Nuclear

    2006-01-01

    Design and safety analyses are presented for conversion of the Portuguese Research Reactor (RPI) from the use of HEU fuel to the use of LEU fuel. The analyses were performed jointly by the RERTR Program at the Argonne National Laboratory (ANL) and the Instituto Tecnologico e Nuclear (ITN). The LEU fuel assembly design uses U{sub 3}Si{sub 2}-Al dispersion fuel with 4.8 g U/cm{sup 3} and is very similar to the HEU fuel design. The results of neutronic studies, steady-state thermal-hydraulic analyses, accident analyses, and revisions to the Operating Limits and Conditions demonstrate that the RPI reactor can be operated safely with the new LEU fuel assemblies. Delivery of the LEU fuel is expected around the end of 2006, with conversion in early 2007. The HEU fuel is planned to be returned to the US in 2008.

  10. The design and performance of the research reactor fuel counter

    SciTech Connect

    Abhold, M.E.; Hsue, S.T.; Menlove, H.O.; Walton, G.; Holt, S.

    1996-09-01

    This paper describes the design features, hardware specifications, and performance characteristics of the Research Reactor Fuel Counter (RRFC) System. The system is an active mode neutron coincidence counter intended to assay material test reactor fuel assemblies under water. The RRFC contains 12 {sup 3}He tubes, each with its own preamplifier, and a single ion chamber. The neutron counting electronics are based on the Los Alamos Portable Shift Register (PSR) and the gamma readout is a manual-range pico-ammeter of Los Alamos design. The RRFC is connected to the surface by a 20-m-long cable bundle. The PSR is controlled by a portable IBM computer running a modified version of the Los Alamos neutron coincidence counting code also called RRFC. There is a manual that describes the RRFC software.

  11. Status of reactor shielding research in the United States

    SciTech Connect

    Bartine, D.E.

    1983-01-01

    Shielding research in the United States continues to place emphasis on: (1) the development and refinement of shielding design calculational methods and nuclear data; and (2) the performance of confirmation experiments, both to evaluate specific design concepts and to verify specific calculational techniques and input data. The successful prediction of the radiation levels observed within the now-operating Fast Flux Test Facility (FFTF) has demonstrated the validity of this two-pronged approach, which has since been applied to US fast breeder reactor programs and is now being used to determine radiation levels and possible further shielding needs at operating light water reactors, especially under accident conditions. A similar approach is being applied to the back end of the fission fuel cycle to verify that radiation doses at fuel element storage and transportation facilities and within fuel reprocessing plants are kept at acceptable levels without undue economic penalties.

  12. Personal neutron dosimetry at a research reactor facility.

    PubMed

    Kamenopoulou, V; Carinou, E; Stamatelatos, I E

    2001-01-01

    Individual neutron monitoring presents several difficulties due to the differences in energy response of the dosemeters. In the present study, an individual dosemeter (TLD) calibration approach is attempted for the personnel of a research reactor facility. The neutron energy response function of the dosemeter was derived using the MCNP code. The results were verified by measurements to three different neutron spectra and were found to be in good agreement. Three different calibration curves were defined for thermal, intermediate and fast neutrons. At the different working positions around the reactor, neutron spectra were defined using the Monte Carlo technique and ambient dose rate measurements were performed. An estimation of the neutrons energy is provided by the ratio of the different TLD pellets of each dosemeter in combination with the information concerning the worker's position; then the dose equivalent is deduced according to the appropriate calibration curve. PMID:11586728

  13. Remote dismantlement activities for the Argonne CP-5 Research Reactor

    SciTech Connect

    Noakes, M.W.

    1996-12-31

    The Department of Energy`s (DOE`s) Robotics Technology Development Program (RTDP) is participating in the dismantlement of a mothballed research reactor, Chicago Pile Number 5 (CP-5), at Argonne National Laboratory (ANL) to demonstrate technology developed by the program while assisting Argonne with their remote system needs. Equipment deployed for CP-5 activities includes the dual-arm work platform (DAWP), which will handle disassembly of reactor internals, and the RedZone Robotics-developed `Rosie` remote work vehicle, which will perform size reduction of shield plugs, demolition of the biological shield, and waste packaging. Remote dismantlement tasks are scheduled to begin in February of 1997 and to continue through 1997 and beyond.

  14. Advanced sodium fast reactor accident source terms : research needs.

    SciTech Connect

    Powers, Dana Auburn; Clement, Bernard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  15. Radioisotope research, production, and processing at the University of Missouri Research Reactor

    SciTech Connect

    Ehrhardt, G.J.; Ketring, A.R.; Ja, Wei; Ma, D.; Zinn, K.; Lanigan, J.

    1995-12-31

    The University of Missouri Research Reactor (MURR) is a 10 MW, light-water-cooled and moderated research reactor which first achieved criticality in 1996 and is currently the highest powered university-owned research reactor in the U.S. For many years a major supplier of reactor-produced isotopes for research and commercial purposes, in the last 15 years MURR has concentrated on development of reactor-produced beta-particle emitters for experimental use in nuclear medicine therapy of cancer and rheumatoid arthritis. MURR has played a major role in the development of bone cancer pain palliation with the agents {sup 153}Sm EDTMP and {sup 186}Re/{sup 188}Re HEDP, as well as in the use of {sup 186}Re, {sup 177}Lu, {sup 166}Ho, and {sup 105}Rh for radioimmunotherapy and receptor-agent-guided radiotherapy. MURR is also responsible for the development of therapeutic, {sup 90}Y-labeled glass microspheres for the treatment of liver tumors, a product ({sup 90}Y Therasphere{trademark}) which is currently an approved drug in Canada. MURR has also pioneered the development of {sup 188}W/{sup 188}Re and {sup 99}Mo/{sup 99m}Tc gel generators, which make the use of low specific activity {sup 188}W and {sup 99}Mo practical for such isotope generators.

  16. High flux research reactors based on particulate fuel

    SciTech Connect

    Powell, J.R.; Takahashi, H.; Horn, F.L.

    1986-02-01

    High Flux Particle Bed Reactor (HFPBR) designs based on High Temperature Gas Reactors (HTGR) particular fuel are described. The coated fuel particles, approx.500 microns in diameter, are packed between porous metal frits, and directly cooled by flowing D/sub 2/O. The large heat transfer surface area in the packed bed, approx.100 cm/sup 2//cm/sup 3/ of volume, allows high power densities, typically 10 MW/liter. Peak thermal fluxes in the HFPBR are 1 to 2 x 1/sup 16/ n/c/sup 2/ sec., depending on configuration and moderator choice with beryllium and D/sub 2/O Moderators yielding the best flux performance. Spent fuel particles can be hydraulically unloaded every day or two and fresh fuel reloaded. The short fuel cycle allows HFPBR fuel loading to be very low, approx.2 kg of /sup 235/U, with a fission product inventory one-tenth of that in present high flux research reactors. The HFPBR can use partially enriched fuel, 20% /sup 235/U, without degradation in flux reactivity. 8 refs., 12 figs., 2 tabs.

  17. Advanced Reactor Safety Research Division. Quarterly progress report, January 1-March 31, 1980

    SciTech Connect

    Agrawal, A.K.; Cerbone, R.J.; Sastre, C.

    1980-06-01

    The Advanced Reactor Safety Research Programs quarterly progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR Safety Evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  18. Present status of liquid metal research for a fusion reactor

    NASA Astrophysics Data System (ADS)

    Tabarés, Francisco L.

    2016-01-01

    Although the use of solid materials as targets of divertor plasmas in magnetic fusion research is accepted as the standard solution for the very challenging issue of power and particle handling in a fusion reactor, a generalized feeling that the present options chosen for ITER will not represent the best choice for a reactor is growing up. The problems found for tungsten, the present selection for the divertor target of ITER, in laboratory tests and in hot plasma fusion devices suggest so. Even in the absence of the strong neutron irradiation expected in a reactor, issues like surface melting, droplet ejection, surface cracking, dust generation, etc., call for alternative solutions in a long pulse, high efficient fusion energy-producing continuous machine. Fortunately enough, decades of research on plasma facing materials based on liquid metals (LMs) have produced a wealth of appealing ideas that could find practical application in the route to the realization of a commercial fusion power plant. The options presently available, although in a different degree of maturity, range from full coverage of the inner wall of the device with liquid metals, so that power and particle exhaust together with neutron shielding could be provided, to more conservative combinations of liquid metal films and conventional solid targets basically representing a sort of high performance, evaporative coating for the alleviation of the surface degradation issues found so far. In this work, an updated review of worldwide activities on LM research is presented, together with some open issues still remaining and some proposals based on simple physical considerations leading to the optimization of the most conservative alternatives.

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

    SciTech Connect

    Clayton, Dwight; Smith, Cyrus

    2014-02-18

    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 and 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.

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

    NASA Astrophysics Data System (ADS)

    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.

  1. Role of research reactors in training of NPP personnel with special focus on training reactor VR-1

    SciTech Connect

    Sklenka, L.; Rataj, J.; Frybort, J.; Huml, O.

    2012-07-01

    Research reactors play an important role in providing key personnel of nuclear power plants a hands-on experience from operation and experiments at nuclear facilities. Training of NPP (Nuclear Power Plant) staff is usually deeply theoretical with an extensive utilisation of simulators and computer visualisation. But a direct sensing of the reactor response to various actions can only improve the personnel awareness of important aspects of reactor operation. Training Reactor VR-1 and its utilization for training of NPP operators and other professionals from Czech Republic and Slovakia is described. Typical experimental exercises and good practices in organization of a training program are demonstrated. (authors)

  2. Oak Ridge National Laboratory Research Reactor Experimenters' Guide

    SciTech Connect

    Cagle, C.D.

    1982-10-01

    The Oak Ridge National Laboratory has three multipurpose research reactors which accommodate testing loops, target irradiations, and beam-type experiments. Since the experiments must share common or similar facilities and utilities, be designed and fabricated by the same groups, and meet the same safety criteria, certain standards for these have been developed. These standards deal only with those properties from which safety and economy of time and money can be maximized and do not relate to the intent of the experiment or quality of the data obtained. The necessity for, and the limitations of, the standards are discussed; and a compilation of general standards is included.

  3. Eastern Europe Research Reactor Initiative nuclear education and training courses - Current activities and future challenges

    SciTech Connect

    Snoj, L.; Sklenka, L.; Rataj, J.; Boeck, H.

    2012-07-01

    The Eastern Europe Research Reactor Initiative was established in January 2008 to enhance cooperation between the Research Reactors in Eastern Europe. It covers three areas of research reactor utilisation: irradiation of materials and fuel, radioisotope production, neutron beam experiments, education and training. In the field of education and training an EERRI training course was developed. The training programme has been elaborated with the purpose to assist IAEA Member States, which consider building a research reactor (RR) as a first step to develop nuclear competence and infrastructure in the Country. The major strength of the reactor is utilisation of three different research reactors and a lot of practical exercises. Due to high level of adaptability, the course can be tailored to specific needs of institutions with limited or no access to research reactors. (authors)

  4. Radiation dosimetry at the BNL Medical Research Reactor

    SciTech Connect

    Holden, N.E.; Reciniello, R.N.; Greenberg, D.D.; Hu, J.P.

    1998-11-01

    The Medical Research Reactor, BMRR, at the Brookhaven National Laboratory, BNL, is a three megawatt, 3 MW, heterogeneous, tank-type, light water cooled and moderated, graphite reflected reactor, which was designed for biomedical studies, and became operational in 1959. It provides thermal and epithermal neutron beams suitable for research studies such as radiation therapy of various types of tumors. At the present time, the major program at BMRR is Boron Neutron Capture Therapy, BNCT. Modifications have been made to the BMRR to significantly increase the available epithermal neutron flux density to a patient in clinical trials of BNCT. The data indicate that the flux density and dose rate are concentrated in the center of the beam, the patient absorbs neutrons rather than gamma radiation and as noted previously even with the increasing flux values, gamma-ray dose received by the attending personnel has remained minimal. Flux densities in the center of the thermal port and epithermal port beams have been characterized with an agreement between the measurements and the calculations.

  5. Unique educational opportunities at the Missouri University research reactor

    SciTech Connect

    Ketring, A.R.; Ross, F.K.; Spate, V.

    1997-12-01

    Since the Missouri University Research Reactor (MURR) went critical in 1966, it has been a center where students from many departments conduct their graduate research. In the past three decades, hundreds of graduate students from the MU departments of chemistry, physics, anthropology, nuclear engineering, etc., have received masters and doctoral degrees based on research using neutrons produced at MURR. More recently, the educational opportunities at MURR have been expanded to include undergraduate students and local high school students. Since 1989 MURR has participated in the National Science Foundation-funded Research Experience for Undergraduates (REU) program. As part of this program, undergraduate students from universities and colleges throughout the United States come to MURR and get hands-on research experience during the summer. Another program, started in 1994 by the Nuclear Analysis Program at MURR, allows students from a local high school to conduct a neutron activation analysis (NAA) experiment. We also conduct tours of the center, where we describe the research and educational programs at MURR to groups of elementary school children, high school science teachers, state legislators, professional organizations, and many other groups.

  6. Nuclear reactor safety research since three mile island.

    PubMed

    Mynatt, F R

    1982-04-01

    The Three Mile Island nuclear power plant accident has resulted in redirection of reactor safety research priorities. The small release to the environment of radioactive iodine-13 to 17 curies in a total radioactivity release of 2.4 million to 13 million curies-has led to a new emphasis on the physical chemistry of fission product behavior in accidents; the fact that the nuclear core was severely damaged but did not melt down has opened a new accident regime-that of the degraded core; the role of the operators in the progression and severity of the accident has shifted emphasis from equipment reliability to human reliability. As research progresses in these areas, the technical base for regulation and risk analysis will change substantially. PMID:17736229

  7. Modular Pebble Bed Reactor Project, University Research Consortium Annual Report

    SciTech Connect

    Petti, David Andrew

    2000-07-01

    This project is developing a fundamental conceptual design for a gas-cooled, modular, pebble bed reactor. Key technology areas associated with this design are being investigated which intend to address issues concerning fuel performance, safety, core neutronics and proliferation resistance, economics and waste disposal. Research has been initiated in the following areas: · Improved fuel particle performance · Reactor physics · Economics · Proliferation resistance · Power conversion system modeling · Safety analysis · Regulatory and licensing strategy Recent accomplishments include: · Developed four conceptual models for fuel particle failures that are currently being evaluated by a series of ABAQUS analyses. Analytical fits to the results are being performed over a range of important parameters using statistical/factorial tools. The fits will be used in a Monte Carlo fuel performance code, which is under development. · A fracture mechanics approach has been used to develop a failure probability model for the fuel particle, which has resulted in significant improvement over earlier models. · Investigation of fuel particle physio-chemical behavior has been initiated which includes the development of a fission gas release model, particle temperature distributions, internal particle pressure, migration of fission products, and chemical attack of fuel particle layers. · A balance of plant, steady-state thermal hydraulics model has been developed to represent all major components of a MPBR. Component models are being refined to accurately reflect transient performance. · A comparison between air and helium for use in the energy-conversion cycle of the MPBR has been completed and formed the basis of a master’s degree thesis. · Safety issues associated with air ingress are being evaluated. · Post shutdown, reactor heat removal characteristics are being evaluated by the Heating-7 code. · PEBBED, a fast deterministic neutronic code package suitable for

  8. Reprocessing of research reactor fuel the Dounreay option

    SciTech Connect

    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.

  9. Monochromatic neutron beam production at Brazilian nuclear research reactors

    NASA Astrophysics Data System (ADS)

    Stasiulevicius, Roberto; Rodrigues, Claudio; Parente, Carlos B. R.; Voi, Dante L.; Rogers, John D.

    2000-12-01

    Monochomatic beams of neutrons are obtained form a nuclear reactor polychromatic beam by the diffraction process, suing a single crystal energy selector. In Brazil, two nuclear research reactors, the swimming pool model IEA-R1 and the Argonaut type IEN-R1 have been used to carry out measurements with this technique. Neutron spectra have been measured using crystal spectrometers installed on the main beam lines of each reactor. The performance of conventional- artificial and natural selected crystals has been verified by the multipurpose neutron diffractometers installed at IEA-R1 and simple crystal spectrometer in operator at IEN- R1. A practical figure of merit formula was introduced to evaluate the performance and relative reflectivity of the selected planes of a single crystal. The total of 16 natural crystals were selected for use in the neutron monochromator, including a total of 24 families of planes. Twelve of these natural crystal types and respective best family of planes were measured directly with the multipurpose neutron diffractometers. The neutron spectrometer installed at IEN- R1 was used to confirm test results of the better specimens. The usually conventional-artificial crystal spacing distance range is limited to 3.4 angstrom. The interplane distance range has now been increased to approximately 10 angstrom by use of naturally occurring crystals. The neutron diffraction technique with conventional and natural crystals for energy selection and filtering can be utilized to obtain monochromatic sub and thermal neutrons with energies in the range of 0.001 to 10 eV. The thermal neutron is considered a good tool or probe for general applications in various fields, such as condensed matter, chemistry, biology, industrial applications and others.

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

    SciTech Connect

    Rosenbalm, K.F.

    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.

  11. Predicting Activation of Experiments Inside the Annular Core Research Reactor

    SciTech Connect

    Greenberg, Joseph Isaac

    2015-11-01

    The objective of this thesis is to create a program to quickly estimate the radioactivity and decay of experiments conducted inside of the Annular Core Research Reactor at Sandia National Laboratories and eliminate the need for users to write code. This is achieved by model the neutron fluxes in the reactor’s central cavity where experiments are conducted for 4 different neutron spectra using MCNP. The desired neutron spectrum, experiment material composition, and reactor power level are then input into CINDER2008 burnup code to obtain activation and decay information for every isotope generated. DREAD creates all of the files required for CINDER2008 through user selected inputs in a graphical user interface and executes the program for the user and displays the resulting estimation for dose rate at various distances. The DREAD program was validated by weighing and measuring various experiments in the different spectra and then collecting dose rate information after they were irradiated and comparing it to the dose rates that DREAD predicted. The program provides results with an average of 17% higher estimates than the actual values and takes seconds to execute.

  12. A research reactor simulator for operators training and teaching

    SciTech Connect

    De Carvalho, R. P.; Maiorino, J. R.

    2006-07-01

    This work describes a training simulator of Research Reactors (RR). The simulator is an interactive tool for teaching and operator training of the bases of the RR operation, reactor physics and thermal hydraulics. The Brazilian IEA-R1 RR was taken as the reference (default configuration). The implementation of the simulator consists of the modeling of the process and system (neutronics, thermal hydraulics), its numerical solution, and the implementation of the man-machine interface through visual interactive screens. The point kinetics model was used for the nuclear process and the heat and mass conservation models were used for the thermal hydraulic feed back in the average core channel. The heat exchanger and cooling tower were also modeled. The main systems were: the reactivity control system, including the automatic control, and the primary and secondary coolant systems. The Visual C++ was used to codes and graphics lay-outs. The simulator is to be used in a PC with Windows XP system. The simulator allows simulation in real time of start up, power maneuver, and shut down. (authors)

  13. The neutron texture diffractometer at the China Advanced Research Reactor

    NASA Astrophysics Data System (ADS)

    Mei-Juan, Li; Xiao-Long, Liu; Yun-Tao, Liu; Geng-Fang, Tian; Jian-Bo, Gao; Zhou-Xiang, Yu; Yu-Qing, Li; Li-Qi, Wu; Lin-Feng, Yang; Kai, Sun; Hong-Li, Wang; R. Santisteban, J.; Dong-Feng, Chen

    2016-03-01

    The first neutron texture diffractometer in China has been built at the China Advanced Research Reactor, due to strong demand for texture measurement with neutrons from the domestic user community. This neutron texture diffractometer has high neutron intensity, moderate resolution and is mainly applied to study texture in commonly used industrial materials and engineering components. In this paper, the design and characteristics of this instrument are described. The results for calibration with neutrons and quantitative texture analysis of zirconium alloy plate are presented. The comparison of texture measurements with the results obtained in HIPPO at LANSCE and Kowari at ANSTO illustrates the reliability of the texture diffractometer. Supported by National Nature Science Foundation of China (11105231, 11205248, 51327902) and International Atomic Energy Agency-TC program (CPR0012)

  14. MCNP-model for the OAEP Thai Research Reactor

    SciTech Connect

    Gallmeier, F.X.; Tang, J.S.; Primm, R.T. III

    1998-06-01

    An MCNP input was prepared for the Thai Research Reactor, making extensive use of the MCNP geometry`s lattice feature that allows a flexible and easy rearrangement of the core components and the adjustment of the control elements. The geometry was checked for overdefined or undefined zones by two-dimensional plots of cuts through the core configuration with the MCNP geometry plotting capabilities, and by a three-dimensional view of the core configuration with the SABRINA code. Cross sections were defined for a hypothetical core of 67 standard fuel elements and 38 low-enriched uranium fuel elements--all filled with fresh fuel. Three test calculations were performed with the MCNP4B-code to obtain the multiplication factor for the cases with control elements fully inserted, fully withdrawn, and at a working position.

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

    SciTech Connect

    Denman, Matthew R.; Porter, Douglas; Wright, Art; Lambert, John; Hayes, Steven; Natesan, Ken; Ott, Larry J.; Garner, Frank; Walters, Leon; Yacout, Abdellatif

    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.

  16. Nuclear plant-aging research on reactor protection systems

    SciTech Connect

    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.

  17. Radiopharmaceuticals developed at the University of Missouri research reactor

    SciTech Connect

    Ketring, A.R.; Ehrhardt, G.J.; Day, D.E.

    1997-12-01

    The University of Missouri Research Reactor (MURR) has put a great deal of effort in the last two decades into development of radiotherapeutic beta emitters as nuclear medicine radiotherapeutics for malignancies. This paper describes the development of two of these drugs, {sup 153}Sm ethylenediaminetetra-methylene phosphonic acid (EDTMP) (Quadramet{trademark}) and {sup 90}Y glass microspheres (TheraSphere{trademark}). Samarium-153 EDTMP is a palliative used to treat the pain of metastatic bone cancer without the side effects of narcotic pain killers. Yttrium-90 glass microspheres are delivered via hepatic artery catheter to embolize the capillaries of liver tumors and deliver a large radiation dose for symptom palliation and life prolonging purposes.

  18. 78 FR 46618 - Order Prohibiting Operation of Aerotest Radiography and Research Reactor

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-08-01

    ... COMMISSION Order Prohibiting Operation of Aerotest Radiography and Research Reactor I. Aerotest Operations... Radiography and Research Reactor (ARRR) in accordance with the conditions specified therein. The ARRR is... and Research and Development,'' of the AEA and 10 CFR 50.38, ``Ineligibility of Certain...

  19. 75 FR 57080 - In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-17

    ... which authorizes the possession, use, and operation of the Aerotest Radiography and Research Reactor... COMMISSION In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order... Regulations (10 CFR) Section 50.21(c) for research and development purposes. Aerotest is a wholly...

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

    SciTech Connect

    Sofu, Tanju; LaChance, Jeffrey L.; Bari, R.; Wigeland, Roald; Denman, Matthew R.; Flanagan, George F.

    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.

  1. Completed Decommissioning of the Research Reactor TRIGA Heidelberg We are specialised in Decommissioning a Research Reactor in Germany now

    SciTech Connect

    Juenger-Graef, B.; Hoever, K.; Moser, T.; Berthold, M.; Blenski, H.J.

    2006-07-01

    This paper describes the decommissioning of the TRIGA Heidelberg II reactor which was used until 1999, and of the TRIGA Heidelberg I reactor, which was for the last 20 years in a safe containment. (authors)

  2. TRIGA reactor facility at the Armed Forces Radiobiology Research Institute: A simplified technical description. revision. Technical report

    SciTech Connect

    Moore, M.L.

    1994-01-01

    This publication provides a simplified technical description of the TRIGA research reactor at AFRRI. Topics covered include general principles of reactor operation and a description of the TRIGA reactor and its unique features.

  3. Second generation Research Reactor Fuel Container (RRFC-II).

    SciTech Connect

    Abhold, M. E.; Baker, M. C.; Bourret, S. C.; Harker, W. C.; Pelowitz, D. G.; Polk, P. J.

    2001-01-01

    The second generation Research Reactor Fuel Counter (RRFC-II) has been developed to measure the remaining {sup 235}U content in foreign spent Material Test Reactor (MTR)-type fuel being returned to the Westinghouse Savannah River Site (WSRS) for interim storage and subsequent disposal. The fuel to be measured started as fresh fuel nominally with 93% enriched Uraniuin alloyed with A1 clad in Al. The fuel was irradiated to levels of up to 65% burnup. The RRFC-II, which will be located in the L-Basin spent fuel pool, is intended to assay the {sup 235}U content using a combination of passive neutron coincidence counting, active neutron coincidence counting, and active-multiplicity analysis. Measurements will be done underwater, eliminating the need for costly and hazardous handling operations of spent fuel out of water. The underwater portion of the RRFC-II consists of a watertight stainless steel housing containing neutron and gamma detectors and a scanning active neutron source. The portion of the system that resides above water consists of data-processing electronics; electromechanical drive electronics; a computer to control the operation of the counter, to collect, and to analyze data; and a touch screen interface located at the equipment rack. The RRFC-II is an improved version of the Los Alamos-designed RRFC already installed in the SRS Receipts Basin for Offsite Fuel. The RRFC-II has been fabricated and is scheduled for installation in late FY 2001 pending acceptance testing by Savannah River Site personnel.

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

    SciTech Connect

    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.

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

    SciTech Connect

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

    1982-01-01

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

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

    SciTech Connect

    Kanda, Keiji; Nishihara, Hedeaki; Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo

    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.

  7. Water Reactor Safety Research Division quarterly progress report, January 1-March 31, 1980

    SciTech Connect

    Romano, A.J.

    1980-06-01

    The Water Reactor Safety Research Programs Quarterly Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evaluation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  8. Water Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

    SciTech Connect

    Abuaf, N.; Levine, M.M.; Saha, P.; van Rooyen, D.

    1980-08-01

    The Water Reactor Safety Research Programs quarterly report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evlauation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  9. 75 FR 39985 - In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-07-13

    ..., use and operation of the Aerotest Radiography and Research Reactor (ARRR) located in San Ramon... the Federal Register on May 14, 2010; 75 FR 27368. No hearing requests or written comments were... COMMISSION In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor);...

  10. Modeling of operating history of the research nuclear reactor

    NASA Astrophysics Data System (ADS)

    Naymushin, A.; Chertkov, Yu; Shchurovskaya, M.; Anikin, M.; Lebedev, I.

    2016-06-01

    The results of simulation of the IRT-T reactor operation history from 2012 to 2014 are presented. Calculations are performed using continuous energy Monte Carlo code MCU-PTR. Comparison is made between calculation and experimental data for the critical reactor.

  11. Development and transfer of fuel fabrication and utilization technology for research reactors

    SciTech Connect

    Travelli, A.; Domagala, R.F.; Matos, J.E.; Snelgrove, J.L.

    1982-01-01

    Approximately 300 research reactors supplied with US-enriched uranium are currently in operation in about 40 countries, with a variety of types, sizes, experiment capabilities and applications. Despite the usefulness and popularity of research reactors, relatively few innovations in their core design have been made in the last fifteen years. The main reason can be better understood by reviewing briefly the history of research reactor fuel technology and enrichment levels. Stringent requirements on the enrichment of the uranium to be used in research reactors were considered and a program was launched to assist research reactors in continuing their operation with the new requirements and with minimum penalties. The goal of the new program, the Reduced Enrichment Research and Test Reactor (RERTR) Program, is to develop the technical means to utilize LEU instead of HEU in research reactors without significant penalties in experiment performance, operating costs, reactor modifications, and safety characteristics. This paper reviews briefly the RERTR Program activities with special emphasis on the technology transfer aspects of interest to this conference.

  12. Core conversion of the Portuguese research reactor to LEU fuel

    SciTech Connect

    Marques, J.G.; Ramos, A.R.; Kocher, A.

    2008-07-15

    Core conversion of the Portuguese Research Reactor (RPI) to LEU fuel is being performed within IAEA's Technical Cooperation project POR/4/016, with financial support from the US and Portugal. CERCA was selected as manufacturer of the LEU assemblies by the IAEA after an international call for bids. CERCA provided a comprehensive package to the RPI which included the mechanical verification of the design of the assemblies, their manufacture and arrangements for a joint inspection of the finished assemblies. The LEU fuel assemblies were manufactured within 8 months upon final approval of the design. The safety analyses for the core conversion to LEU fuel were made with the assistance of the RERTR program and were submitted for review by the IAEA and by Portuguese authorities in January 2007. Revised documents were submitted in June 2007 addressing the issues raised during review. Regulatory approval was received in early August and core conversion was done in early September. All measured safety parameters are within the defined acceptance limits. Operation at full power is expected by the end of October. (author)

  13. Characterization of Novel Calorimeters in the Annular Core Research Reactor

    NASA Astrophysics Data System (ADS)

    Hehr, Brian D.; Parma, Edward J.; Peters, Curtis D.; Naranjo, Gerald E.; Luker, S. Michael

    2016-02-01

    A series of pulsed irradiation experiments have been performed in the central cavity of Sandia National Laboratories' Annular Core Research Reactor (ACRR) to characterize the responses of a set of elemental calorimeter materials including Si, Zr, Sn, Ta, W, and Bi. Of particular interest was the perturbing effect of the calorimeter itself on the ambient radiation field - a potential concern in dosimetry applications. By placing the calorimeter package into a neutron-thermalizing lead/polyethylene (LP) bucket and irradiating both with and without a cadmium wrapper, it was demonstrated that prompt capture gammas generated inside the calorimeters can be a significant contributor to the measured dose in the active disc region. An MCNP model of the experimental setup was shown to replicate measured dose responses to within 10%. The internal (n,γ) contribution was found to constitute as much as 50% of the response inside the LP bucket and up to 20% inside the nominal (unmodified) cavity environment, with Ta and W exhibiting the largest enhancement due to their sizable (n,γ) cross sections. Capture reactions in non-disc components of the calorimeter were estimated to be responsible for up to a few percent of the measured response. This work was supported by the United States Department of Energy under Contract DE-AC04-94AL85000. Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy.

  14. Condensed matter research at the modernized IBR-2 reactor: from functional materials to nanobiotechnologies

    NASA Astrophysics Data System (ADS)

    Aksenov, V. L.; Balagurov, A. M.; Kozlenko, D. P.

    2016-07-01

    An overview of the main scientific areas of condensed matter research, which are extended with the use of the IBR-2 high-flux research reactor, is presented. It is demonstrated that the spectrometer facility of the upgraded reactor has great potential for studying the structural, magnetic, and dynamical properties of novel functional materials and nanobiosystems, which ensures the leading position of the Joint Institute for Nuclear Research in neutron research of condensed matter for the long-term prospect.

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

    SciTech Connect

    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.

  16. Characterization of debris/concrete interactions for advanced research reactor and commercial BWR severe accidents

    SciTech Connect

    Hyman, C.R.; Taleyarkhan, R.P.; Greene, S.R.

    1991-01-01

    The core concrete interaction (CCI) is an important phase of any severe accident where the reactor vessel has failed and core debris is relocated onto the containment basemat. In recent calculations performed at the Oak Ridge National Laboratory (ORNL), CCI has been studied for severe accidents occurring in a commercial Boiling Water Reactor (BWR) and in a high-power density Department of Energy (DOE) research reactor that is currently in the conceptual design stage. Because of differences in the debris decay heating level, core debris composition and inventory, and containment design, the characteristics of the resulting CCI and containment response are different for the two reactor types. Furthermore, proper selection of the basemat concrete type and the provision of an overlying water pool are found to be significant CCI mitigating factors for the research reactor and thus constitute important design considerations for any future reactor type. 10 refs., 4 figs., 1 tab.

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

    SciTech Connect

    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.

  18. A new safety channel based on ¹⁷N detection in research reactors.

    PubMed

    Seyfi, Somayye; Gharib, Morteza

    2015-10-01

    Tehran research reactor (TRR) is a representative of pool type research reactors using light water, as coolant and moderator. This reactor is chosen as a prototype to demonstrate and prove the feasibility of (17)N detection as a new redundant channel for reactor power measurement. In TRR, similar to other pool type reactors, neutron detectors are immersed in the pool around the core as the main power measuring devices. In the present article, a different approach, using out of water neutron detector, is employed to measure reactor power. This new method is based on (17)O (n,p) (17)N reaction taking place inside the core and subsequent measurement of delayed neutrons emitted due to (17)N disintegration. Count and measurement of neutrons around outlet water pipe provides a reliable redundant safety channel to measure reactor power. Results compared with other established channels indicate a good agreement and shows a linear interdependency with true thermal power. Safety of reactor operation is improved with installation & use of this new power measuring channel. The new approach may equally serve well as a redundant channel in all other types of reactors having coolant comprised of oxygen in its molecular constituents. Contrary to existing channels, this one is totally out of water and thus is an advantage over current instrumentations. It is proposed to employ the same idea on other reactors (nuclear power plants too) to improve safety criteria. PMID:26123105

  19. Code System to Calculate Mixed Cores in TRIGA Mark II Research Reactor.

    Energy Science and Technology Software Center (ESTSC)

    2001-08-29

    Version 00 TRIGLAV is a computer program for reactor calculations of mixed cores in a TRIGA Mark II research reactor. It can be applied for fuel element burn-up calculations, for power and flux distributions calculations and for reactivity predictions. The TRIGLAV program requires the WIMS-D4 program with the original WIMS cross-section library extended for TRIGA reactor specific nuclides. This package includes the code TRIGAC, which is a new version of TRIGAP.

  20. Status of reactor-shielding research in the US

    SciTech Connect

    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.

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

    SciTech Connect

    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.

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

    SciTech Connect

    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.

  3. Thermonuclear Fusion Research Progress and the Way to the Reactor

    NASA Astrophysics Data System (ADS)

    Koch, Raymond

    2006-06-01

    The paper reviews the progress of fusion research and its prospects for electricity generation. It starts with a reminder of the principles of thermonuclear fusion and a brief discussion of its potential role in the future of the world energy production. The reactions allowing energy production by fusion of nuclei in stars and on earth and the conditions required to sustain them are reviewed. At the high temperatures required for fusion (hundred millions kelvins), matter is completely ionized and has reached what is called its 4th state: the plasma state. The possible means to achieve these extreme temperatures is discussed. The remainder of the paper focuses on the most promising of these approaches, magnetic confinement. The operating principles of the presently most efficient machine of this type — the tokamak — is described in some detail. On the road to producing energy with fusion, a number of obstacles have to be overcome. The plasma, a fluid that reacts to electromagnetic forces and carries currents and charges, is a complex medium. Fusion plasma is strongly heated and is therefore a good example of a system far from equilibrium. A wide variety of instabilities can grow in this system and lead to self-organized structures and spontaneous cycles. Turbulence is generated that degrades the confinement and hinders easy achievement of long lasting hot plasmas. Physicists have learned how to quench turbulence, thereby creating sort of insulating bottles inside the plasma itself to circumvent this problem. The recent history of fusion performance is outlined and the prospect of achieving power generation by fusion in a near future is discussed in the light of the development of the "International Tokamak Experimental Reactor" project ITER.

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

    NASA Astrophysics Data System (ADS)

    Koerner, S.; Schillinger, B.; Vontobel, P.; Rauch, H.

    2001-09-01

    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 this beam position is 1.3×10 5 neutrons/cm 2 s and the beam diameter is 8 cm. For a 3D tomographic reconstruction of the sample interior, transmission images of the object taken from different view angles are required. Therefore, a rotary table driven by a step motor connected to a computerized motion control system has been installed at the sample position. In parallel a suitable electronic imaging device based on a neutron sensitive scintillator screen and a CCD-camera has been designed. It can be controlled by a computer in order to synchronize the software of the detector and of the rotary table with the aim of an automation of measurements. Reasonable exposure times can get as low as 20 s per image. This means that a complete tomography of a sample can be performed within one working day. Calculation of the 3D voxel array is made by using the filtered backprojection algorithm.

  5. Gas release driven dynamics in research reactors piping

    SciTech Connect

    Kolev, Nikolay Ivanov; Roloff-Bock, Iris; Schlicht, Gerhard

    2006-07-01

    Analysis of the physical and chemical processes of radiolysis gas production, air absorption, diffusion controlled gas release and transport in the coolant cleaning system of the research reactor FRM II, which is now being in routine power operation in Munich, Germany, lead to the following conclusions: 1) The steady state pressure distribution in the siphon pipe allows that the horizontal part of the siphon pipe is filled with air. The air is isolated by about 1 m water column from the main pipe of the coolant cleaning system (CCS). This is a stable steady state. It has two positive impacts on the normal operation of the CCS: (a) there is effectively no bypass flow; (b) The air can not be transported through the pipe and therefore no deterioration of the pump performance is expected from the function of the siphon pipe. 2) Radiolysis gas production for coolant, that initially does not contain dissolved air, does not lead to any problem for the system. The gases are dissolved in the coolant at 2.2 bar and are not released for pressures reduction to about 1 bar, which is the minimum pressure in the CCS. 3) Assuming hypothetically a radiolysis gas production for coolant, which initially does contain dissolved air close to its saturation, leads to gas slug formation and its transport up to the pump. This could reduce the pump head and could lead to distortion of the normal operation. Systematic measurement of the hydrogen in the primary system at 100% power indicated, that this state is not realized in the system. The observed H{sub 2} concentration was between 0.016 e-6 and 0.380 e-6 which is of no concern at all. (authors)

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

    SciTech Connect

    Smith, Cyrus M; Nanstad, Randy K; Clayton, Dwight A; Matlack, Katie; Ramuhalli, Pradeep; Light, Glenn

    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.

  7. Thermal-hydraulic aspects of flow inversion in a research reactor

    SciTech Connect

    Smith, R.S.; Woodruff, W.L.

    1986-11-01

    PARET, a neutronics and thermal-hydraulics computer code, has been modified to account for natural convection in a reactor core. The code was then used to analyze the flow inversion that occurs in a reactor with heat removal by forced convection in the downward direction after a pump failure. Typical results are shown for a number of parameters. Research reactors normally operating much above ten MW are predicted to experience nucleate boiling in the event of a flow inversion. Comparison with experimental results from the Belgian BR2 reactor indicated general agreement although nucleate boiling that was analytically predicted was not noted in the BR2 data.

  8. Boron neutron capture therapy and radiation synovectomy research at the Massachusetts Institute of Technology Research Reactor

    SciTech Connect

    Zamenhof, R.G.; Nwanguma, C.I.; Wazer, D.E.; Saris, S.; Madoc-Jones, H. ); Sledge, C.B.; Shortkroff, S. )

    1992-04-01

    In this paper, current research in boron neutron capture therapy (BNCT) and radiation synovectomy at the Massachusetts Institute of Technology Research Reactor is reviewed. In the last few years, major emphasis has been placed on the development of BNCT primarily for treatment of brain tumors. This has required a concerted effort in epithermal beam design and construction as well as the development of analytical capabilities for {sup 10}B analysis and patient treatment planning. Prompt gamma analysis and high-resolution track-etch autoradiography have been developed to meet the needs, respectively, for accurate bulk analysis and for quantitative imaging of {sup 10}B in tissue at subcellular resolutions. Monte Carlo-based treatment planning codes have been developed to ensure optimized and individualized patient treatments. In addition, the development of radiation synovectomy as an alternative therapy to surgical intervention is joints that are affected by rheumatoid arthritis is described.

  9. Efficient testing of ITER materials and components at the Research Institute of Atomic Reactors` experimental facilities

    SciTech Connect

    Ivanov, V.; Kazakov, V.; Pokrovsky, A.; Shamardin, V.; Melder, R.; Revyakin, Yu.; Sandakov, V.

    1995-12-31

    The Research Institute of Atomic Reactors (RIAR) of the State Scientific Centre of the Russian Federation has carried out reactor tests of fusion reactor materials and components. RIAR contains an ideal complex of installations, experimental setups, and diagnostics for such investigations. It includes several different types of reactors, including a fast neutron reactor, a high-flux intermediate-neutron SM-3 reactor, a intermediate-neutron loop reactor, and two RBT-type reactors, and a hot cells complex with remote handling facilities to allow study of the physical-mechanical properties, structure, and elemental composition of irradiated materials. RIAR has carried out a number of initial experiments, including testing of copper and vanadium alloys, electro-insulative coatings, steels, ceramics, diagnostic systems materials, and in-core and hot cell set-ups for divertor mock-up testing, and has collaborative efforts underway with the Scientific Research Institute Electrophysical Apparatus-St. Petersburg (SRIEA), Oak Ridge National Laboratory (ORNL), Argonne National Laboratory (ANL), Red Star, the Institute of Physics and Power engineering (IPPE), the Scientific Research Institute of Inorganic Materials (SRIIM), and Pacific Northwest Laboratory (PNL).

  10. Concept of an accelerator-driven subcritical research reactor within the TESLA accelerator installation

    NASA Astrophysics Data System (ADS)

    Pešić, Milan; Nešković, Nebojša

    2006-06-01

    Study of a small accelerator-driven subcritical research reactor in the Vinča Institute of Nuclear Sciences was initiated in 1999. The idea was to extract a beam of medium-energy protons or deuterons from the TESLA accelerator installation, and to transport and inject it into the reactor. The reactor core was to be composed of the highly enriched uranium fuel elements. The reactor was designated as ADSRR-H. Since the use of this type of fuel elements was not recommended any more, the study of a small accelerator-driven subcritical research reactor employing the low-enriched uranium fuel elements began in 2004. The reactor was designated as ADSRR-L. We compare here the results of the initial computer simulations of ADSRR-H and ADSRR-L. The results have confirmed that our concept could be the basis for designing and construction of a low neutron flux model of the proposed accelerator-driven subcritical power reactor to be moderated and cooled by lead. Our objective is to study the physics and technologies necessary to design and construct ADSRR-L. The reactor would be used for development of nuclear techniques and technologies, and for basic and applied research in neutron physics, metrology, radiation protection and radiobiology.

  11. Integrated Decision-Making Tool to Develop Spent Fuel Strategies for Research Reactors

    SciTech Connect

    Beatty, Randy L; Harrison, Thomas J

    2016-01-01

    IAEA Member States operating or having previously operated a Research Reactor are responsible for the safe and sustainable management and disposal of associated radioactive waste, including research reactor spent nuclear fuel (RRSNF). This includes the safe disposal of RRSNF or the corresponding equivalent waste returned after spent fuel reprocessing. One key challenge to developing general recommendations lies in the diversity of spent fuel types, locations and national/regional circumstances rather than mass or volume alone. This is especially true given that RRSNF inventories are relatively small, and research reactors are rarely operated at a high power level or duration typical of commercial power plants. Presently, many countries lack an effective long-term policy for managing RRSNF. This paper presents results of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) #T33001 on Options and Technologies for Managing the Back End of the Research Reactor Nuclear Fuel Cycle which includes an Integrated Decision Making Tool called BRIDE (Back-end Research reactor Integrated Decision Evaluation). This is a multi-attribute decision-making tool that combines the Total Estimated Cost of each life-cycle scenario with Non-economic factors such as public acceptance, technical maturity etc and ranks optional back-end scenarios specific to member states situations in order to develop a specific member state strategic plan with a preferred or recommended option for managing spent fuel from Research Reactors.

  12. Effect of reduced enrichment on the fuel cycle for research reactors

    SciTech Connect

    Travelli, A.

    1982-01-01

    The new fuels developed by the RERTR Program and by other international programs for application in research reactors with reduced uranium enrichment (<20% EU) are discussed. It is shown that these fuels, combined with proper fuel-element design and fuel-management strategies, can provide at least the same core residence time as high-enrichment fuels in current use, and can frequently significantly extend it. The effect of enrichment reduction on other components of the research reactor fuel cycle, such as uranium and enrichment requirements, fuel fabrication, fuel shipment, and reprocessing are also briefly discussed with their economic implications. From a systematic comparison of HEU and LEU cores for the same reference research reactor, it is concluded that the new fuels have a potential for reducing the research reactor fuel cycle costs while reducing, at the same time, the uranium enrichment of the fuel.

  13. The development of an on-line ERM system for the research reactors in Korea

    NASA Astrophysics Data System (ADS)

    Kim, Hee Reyoung; Lee, Wanno; Kim, Eun Han; Choi, Geun Sik; Lee, Chang Woo

    2007-08-01

    A real-time on-line environmental radiation monitoring (ERM) system for the research reactor sites of Daejeon and Seoul is established. In the Daejeon site, a radio communication method with a radiofrequency of 468.8 MHz is used between the main computer and the six posts inside the Daejeon research reactor site. A general telephone communication method by a dial modem is applied between the main computer and a comparison point with one post outside the Daejeon research reactor site. In the Seoul site, a null modem communication method is employed between a sub-computer and the three posts inside the Seoul research reactor site, and a high-speed communication network such as ADSL is used between the sub-computer in the Seoul site and the main computer in the Daejeon site. Consequently, the real-time data from a total of 10 places is displayed on-line on a screen and it is statistically treated.

  14. A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

    SciTech Connect

    Emiliano Pozzi; David W. Nigg; Marcelo Miller; Silvia I. Thorp; Amanda E. Schwint; Elisa M. Heber; Veronica A. Trivillin; Leandro Zarza; Guillermo Estryk

    2007-11-01

    The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated.

  15. TRIGA reactor facility at the Armed Forces Radiobiology Research Institute: a simplified technical description. Technical report

    SciTech Connect

    Moore, M.L.; Elsasser, S.

    1986-05-01

    In support of its mission the Armed Forces Radiobiology Research Institute (AFRRI) operates a medium-sized research nuclear reactor. The reactor is used to generate radiations, primarily neutrons and gamma rays, which are used to conduct experimental biomedical research and to produce isotopes. The radiations are delivered to the experiments in one of two ways: a pulse operation delivers a very short burst of high power, or a steady-state operation delivers a longer, continuous low- to medium-power exposure. The reactor is also used to train military personnel in reactor operations. TRIGA is an acronym for Training, Research, and Isotope, General Atomics. Mark-F is the specific General Atomics Reactor model, distinguished by a pool, a movable core, exposure-room facilities, and the ability to pulse to momentary high powers. Reactor operations at AFRRI began is 1962. In 1965, a change was made from aluminum-clad to stainless steel-clad fuel elements. Currently more than 150 multiple exposure experiments are performed each year using the reactor.

  16. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  17. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  18. Sensitivity and uncertainty analyses for thermo-hydraulic calculation of research reactor

    SciTech Connect

    Hartini, Entin; Andiwijayakusuma, Dinan; Isnaeni, Muh Darwis

    2013-09-09

    The sensitivity and uncertainty analysis of input parameters on thermohydraulic calculations for a research reactor has successfully done in this research. The uncertainty analysis was carried out on input parameters for thermohydraulic calculation of sub-channel analysis using Code COOLOD-N. The input parameters include radial peaking factor, the increase bulk coolant temperature, heat flux factor and the increase temperature cladding and fuel meat at research reactor utilizing plate fuel element. The input uncertainty of 1% - 4% were used in nominal power calculation. The bubble detachment parameters were computed for S ratio (the safety margin against the onset of flow instability ratio) which were used to determine safety level in line with the design of 'Reactor Serba Guna-G. A. Siwabessy' (RSG-GA Siwabessy). It was concluded from the calculation results that using the uncertainty input more than 3% was beyond the safety margin of reactor operation.

  19. Sensitivity and uncertainty analyses for thermo-hydraulic calculation of research reactor

    NASA Astrophysics Data System (ADS)

    Hartini, Entin; Andiwijayakusuma, Dinan; Isnaeni, Muh Darwis

    2013-09-01

    The sensitivity and uncertainty analysis of input parameters on thermohydraulic calculations for a research reactor has successfully done in this research. The uncertainty analysis was carried out on input parameters for thermohydraulic calculation of sub-channel analysis using Code COOLOD-N. The input parameters include radial peaking factor, the increase bulk coolant temperature, heat flux factor and the increase temperature cladding and fuel meat at research reactor utilizing plate fuel element. The input uncertainty of 1% - 4% were used in nominal power calculation. The bubble detachment parameters were computed for S ratio (the safety margin against the onset of flow instability ratio) which were used to determine safety level in line with the design of "Reactor Serba Guna-G. A. Siwabessy" (RSG-GA Siwabessy). It was concluded from the calculation results that using the uncertainty input more than 3% was beyond the safety margin of reactor operation.

  20. Failure of TRIGA fuel cladding at the Berkeley research reactor

    SciTech Connect

    Denton, Michael M.; Lim, Tek H.

    1986-07-01

    Following a long maintenance shutdown during which a fission chamber was refurbished and a compensated ion chamber replaced, concentrations of radioisotopes were detected in the reactor-room air on a Constant (CAM) after two and a half hours of full-power operation. Following test lead to identification of three fission-product gasses in the reactor room air: Kr{sup 85m}, Kr{sup 87} , and Kr{sup 88} . Conservative estimates indicated the maximum concentrations of all fission gasses to be about 1.1x10{sup -8} {mu}Ci/ml with a total release of less than 1 mCi. It was concluded that the gasses come from a leaking fuel element. Three old, instrumented elements with defective thermocouples were removed first and the reactor was tested at full-power. No abnormal activities were detected during or following the operation. Each of the suspected fuel elements are instrumented with leadout tubes extending 15 feet to above the pool surface. This suggests some possible causes for the cladding failure. First, flexing due to daily movement of the core could have weakened the tube/cladding connection. Secondly, the cladding itself may have been damaged during maintenance procedures requiring removal of the elements or repositioning of the leadout tubes.

  1. Reactor safety research programs. Quarterly report, July-September 1983

    SciTech Connect

    Edler, S.K.

    1984-04-01

    Evaluations of nondestructive examination (NDE) techniques and instrumentation 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, and examining NDE reliability and probabilistic fracture mechanics. Accelerated pellet-cladding interaction modeling is being conducted to predict the probability of fuel rod failure under normal operating conditions. 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. Experimental data and validated models are being used to determine a method for evaluating the acceptance of welded or weld-repaired stainless steel piping. Thermal-hydraulic models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. High-temperature materials property tests are being conducted to provide data on severe core damage fuel behavior. Severe fuel damage accident tests are being conducted at the NRU reactor, Chalk River, Canada; and an instrumented fuel assembly irradiation program is being performed at Halden, Norway. Fuel assemblies and analytical support are being provided for experimental programs at other facilities, including the Super Sara Test Program, Ispra, Italy, and experimental programs at the Power Burst Facility.

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

    SciTech Connect

    Edler, S.K.

    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.

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

    SciTech Connect

    Edler, S.K.

    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.

  4. Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

    SciTech Connect

    Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.; Severynse, T.F.; Sindelar, R.L.; Moore, E.N.

    2013-07-01

    Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage

  5. PROSPECT Background Studies and Operation of Li-Loaded Liquid Scintillator Detectors at a Research Reactor

    NASA Astrophysics Data System (ADS)

    Langford, Thomas; Prospect Collaboration

    2015-04-01

    Segmented antineutrino detectors placed near compact research reactors provide an excellent opportunity to probe short-baseline neutrino oscillations and precisely measure the reactor antineutrino spectrum. PROSPECT is a phased experiment that will explore the favored reactor anomaly parameter space at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab. Measurements of the reactor correlated and ambient backgrounds will be presented, as well as a discussion of active and passive mitigation plans. A lithium-loaded liquid scintillator test detector is currently in operation at HFIR within a prototype shielding cave. Results from recent operation will be presented along with a discussion of their impact on PROSPECT. on behalf of the PROSPECT collaboration.

  6. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

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

    NASA Astrophysics Data System (ADS)

    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.

  8. Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report

    SciTech Connect

    Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

    2002-11-01

    This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOE’s Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

  9. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

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

    SciTech Connect

    Taylor, K.E.; Holm, R.L.

    2006-07-01

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

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

    PubMed

    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. PMID:21459586

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

    SciTech Connect

    Trosman, H.G.; 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.

  13. Photon spectrum behind biological shielding of the LVR-15 research reactor

    SciTech Connect

    Klupak, V.; Viererbl, L.; Lahodova, Z.; Marek, M.; Vins, M.

    2011-07-01

    The LVR-15 reactor is a light water research reactor situated at the Research Centre Rez, near Prague. It operates as a multipurpose facility with a maximum thermal power of 10 MW. The reactor core usually contains from 28 to 32 fuel assemblies with a total mass of {sup 235}U of about 5 kg. Emitted radiation from the fuel caused by fission is shielded by moderating water, a steel reactor vessel, and heavy concrete. This paper deals with measurement and analysis of the gamma spectrum near the outer surface of the concrete wall, behind biological shielding, mainly in the 3- to 10-MeV energy range. A portable HPGe detector with a portable multichannel analyzer was used to measure gamma spectra. The origin of energy lines in gamma detector spectra was identified. (authors)

  14. Spain-Chile and Spain-Ecuador cooperation in the field of research nuclear reactors

    SciTech Connect

    Avendano, G.; Rodriguez, M.L.; Manas, L.; Masalleras, E.; Montes, J.

    1981-01-01

    The Spanish Board of Nuclear Energy (JEN) has been cooperating for the last several years with the Comision Chilena de Energia Nuclear (Chilean Commission of Nuclear Energy (CCHEN)), on the one hand, and with the Comision Ecuatoriana de Energia Atomica (Ecuadorian Commission of Atomic Energy (CEEA)), on the other. The result of this cooperation has been the implementation of projects in both countries to create research centers around a nuclear reactor as the main working tool: the Lo Aguirre reactor in Chile and the Ruminahui reactor in Ecuador.

  15. Unique applications of research reactors with TRIGA UZrH[sub x] fuel

    SciTech Connect

    Whittemore, W.L. )

    1993-01-01

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

  16. Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel

    SciTech Connect

    Matos, J.E.; Mo, S.C.; Woodruff, W.L.

    1992-01-01

    The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. Results of design and safety analyses performed by the RERTR Program at the Argonne National Laboratory for LEU conversion of the GTRR are summarized. Only those parameters which could change as a result of replacing the fuel are addressed. The performance of the reactor and all safety margins with LEU fuel are expected to be about the same as those with the current HEU fuel.

  17. Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel

    SciTech Connect

    Matos, J.E.; Mo, S.C.; Woodruff, W.L.

    1992-12-31

    The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. Results of design and safety analyses performed by the RERTR Program at the Argonne National Laboratory for LEU conversion of the GTRR are summarized. Only those parameters which could change as a result of replacing the fuel are addressed. The performance of the reactor and all safety margins with LEU fuel are expected to be about the same as those with the current HEU fuel.

  18. A human reliability analysis of the University of New Mexico's AGN- 201M nuclear research reactor

    SciTech Connect

    Brumburgh, G.P. ); Heger, A.S. . Dept. of Chemical and Nuclear Engineering)

    1992-10-15

    During 1990--1991, a probabilistic risk assessment was conducted on the University of New Mexico's AGN-201M nuclear research reactor to address the risk and consequence of a maximum hypothetical release accident. The assessment indicated a potential for consequential human error to precipitate Chis scenario. Subsequently, a human reliability analysis was performed to evaluate the significance of human interaction on the reactor's safety systems. This paper presents the results of that investigation.

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

    SciTech Connect

    Not Available

    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.

  20. Research and development of an electrochemical biocide reactor

    NASA Technical Reports Server (NTRS)

    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.

  1. Research of a Supercritical Pressure Water Cooled Reactor in Korea

    SciTech Connect

    Bae, Yoon-Yeong; Joo, Hyung-Kook; Jang, Jinsung; Jeong, Yong-Hwan; Song, Jin-ho; Yoon, Han-Young; Yoo, Jung-Yul

    2004-07-01

    In this paper the activities on the supercritical pressure water-cooled reactor (SCWR) in Korea are briefly introduced. Four projects on a SCWR are being conducted in Korea. Three of them are supported by the I-NERI program while one is by KAERI. Two of the I-NERI-supported projects concern suitable materials for supercritical pressure and temperature, and radiation environment. The other I-NERI-supported project surveys numerically and experimentally the proper turbulence modeling for the numerical calculation of heat transfer phenomena at a supercritical condition. Heat transfer at a supercritical condition is being studied at KAERI experimentally using carbon dioxide as a coolant. The test loop is to be completed by the end of 2004. (authors)

  2. REACTOR

    DOEpatents

    Spitzer, L. Jr.

    1962-01-01

    The system conteraplates ohmically heating a gas to high temperatures such as are useful in thermonuclear reactors of the stellarator class. To this end the gas is ionized and an electric current is applied to the ionized gas ohmically to heat the gas while the ionized gas is confined to a central portion of a reaction chamber. Additionally, means are provided for pumping impurities from the gas and for further heating the gas. (AEC)

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

    ERIC Educational Resources Information Center

    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…

  4. Using low-enriched uranium in research reactors: The RERTR program

    SciTech Connect

    Travelli, A.

    1994-05-01

    The goal of the RERTR program is to minimize and eventually eliminate use of highway enriched uranium (HEU) in research and test reactors. The program has been very successful, and has developed low-enriched uranium (LEU) fuel materials and designs which can be used effectively in approximately 90 percent of the research and test reactors which used HEU when the program began. This progress would not have been possible without active international cooperation among fuel developers, commercial vendors, and reactor operators. The new tasks which the RERTR program is undertaking at this time include development of new and better fuels that will allow use of LEU fuels in all research and test reactors; cooperation with Russian laboratories, which will make it possible to minimize and eventually eliminate use of HEU in research reactors throughout the world, irrespective of its origin; and development of an LEU-based process for the production of {sup 99}Mo. Continuation and intensification of international cooperation are essential to the achievement of the ultimate goals of the RERTR program.

  5. A kinetics and thermalhydraulics capability for the analysis of research reactors

    SciTech Connect

    Woodruff, W.L.

    1984-02-01

    The PARET code, originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors, has now been modified to include a selection of flow instability, departure from nucleate boiling, single-and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model that estimates the voiding produced by subcooled boiling. This modified code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons have been made with the experimental results from the SPERT-I transients, and the agreement with the experimental data is generally quite good. The selection of proper correlations and properties for the range of interest in research reactors was essential to the accuracy of the results. The code has also been applied to the analysis of the International Atomic Energy Agency 10-MW benchmark cores for protected and unprotected transients. The code provides an accurate capability for the analysis of research reactor transients. This modified version of the PARET code is available through the National Energy Software Center.

  6. A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor.

    SciTech Connect

    Bretscher, M. M.

    1998-10-14

    The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% {sup 235}U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm{sup 3} and 3.8 gU/cm{sup 3} are required to match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively.

  7. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    SciTech Connect

    Bissani, M; O'Kelly, D S

    2006-05-08

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to provide color-enhanced gemstones but is

  8. Los Alamos National Laboratory case studies on decommissioning of research reactors and a small nuclear facility

    SciTech Connect

    Salazar, M.D.

    1998-12-01

    Approximately 200 contaminated surplus structures require decommissioning at Los Alamos National Laboratory. During the last 10 years, 50 of these structures have undergone decommissioning. These facilities vary from experimental research reactors to process/research facilities contaminated with plutonium-enriched uranium, tritium, and high explosives. Three case studies are presented: (1) a filter building contaminated with transuranic radionuclides; (2) a historical water boiler that operated with a uranyl-nitrate solution; and (3) the ultra-high-temperature reactor experiment, which used enriched uranium as fuel.

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

    SciTech Connect

    Not Available

    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.

  10. Utilization of the Philippine Research Reactor as a training facility for nuclear power plant operators

    SciTech Connect

    Palabrica, R.J.

    1981-01-01

    The Philippines has a 1-MW swimming-pool reactor facility operated by the Philippine Atomic Energy Commission (PAEC). The reactor is light-water moderated and cooled, graphite reflected, and fueled with 90% enriched uranium. Since it became critical in 1963 it has been utilized for research, radioisotope production, and training. It was used initially in the training of PAEC personnel and other research institutions and universities. During the last few years, however, it has played a key role in training personnel for the Philippine Nuclear Power Project (PNPP).

  11. KNUDSEN CELL REACTOR FOR CATALYST RESEARCH RELATED TO HYDROGEN TECHNOLOGIES

    EPA Science Inventory

    Hydrogen has been identified as a viable sustainable alternative to fossil fuels. Hydrogen as an energy source is ecologically feasible, socially desirable, and with continued research and development promises to become economically viable. The faculty advisors listed...

  12. Collection and analysis of Health Physics Research Reactor operational and use data

    SciTech Connect

    Sims, C.S.

    1985-04-01

    The Health Physics Research Reactor (HPRR) is the primary research tool at the Dosimetry Applications Research (DOSAR) Facility. In addition to use by the DOSAR staff, the HPRR is used by a wide segment of the scientific community for a variety of experimental purposes. This report is a compilation and analysis of data concerning HPRR uses, users, and operations through the end of FY 1984. 17 refs., 12 tabs.,

  13. Research Reactor Preparations for the Air Shipment of Highly Enriched Uranium from Romania

    SciTech Connect

    K. J. Allen; I. Bolshinsky; L. L. Biro; M. E. Budu; N. V. Zamfir; M. Dragusin; C. Paunoiu; M. Ciocanescu

    2010-03-01

    In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation for conversion to low enriched uranium. The Institute for Nuclear Research at Pitesti (SCN Pitesti) shipped 30.1 kg of HEU fresh fuel pellets to Dimitrovgrad, Russia and the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH) shipped 23.7 kilograms of HEU spent fuel assemblies from the VVR S research reactor at Magurele, Romania, to Chelyabinsk, Russia. Both HEU shipments were coordinated by the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), were managed in Romania by the National Commission for Nuclear Activities Control (CNCAN), and were conducted in cooperation with the Russian Federation State Corporation Rosatom and the International Atomic Energy Agency. Both shipments were transported by truck to and from respective commercial airports in Romania and the Russian Federation and stored at secure nuclear facilities in Russia until the material is converted into low enriched uranium. These shipments resulted in Romania becoming the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the research reactor preparations and license approvals that were necessary to safely and securely complete these air shipments of nuclear fuel.

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

    SciTech Connect

    Talley, Darren G.

    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.

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

    SciTech Connect

    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.

  16. Foreign research reactor irradiated nuclear fuel inventories containing HEU and LEU of United States origin

    SciTech Connect

    Matos, J.E.

    1994-12-01

    This report provides estimates of foreign research reactor inventories of aluminum-based and TRIGA irradiated nuclear fuel elements containing highly enriched and low enriched uranium of United States origin that are anticipated in January 1996, January 2001, and January 2006. These fuels from 104 research reactors in 41 countries are the same aluminum-based and TRIGA fuels that were eligible for receipt under the Department of Energy`s Offsite Fuels Policy that was in effect in 1988. All fuel inventory and reactor data that were available as of December 1, 1994, have been included in the estimates of approximately 14,300 irradiated fuel elements in January 1996, 18,800 in January 2001, and 22,700 in January 2006.

  17. Characterization of the Annular Core Research Reactor (ACRR) Neutron Radiography System Imaging Plane

    NASA Astrophysics Data System (ADS)

    Kaiser, Krista; Chantel Nowlen, K.; DePriest, K. Russell

    2016-02-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1) available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  18. Radiation dosimetry for NCT facilities at the Brookhaven Medical Research Reactor

    SciTech Connect

    Holden, N.E.; Hu, J.P.; Greenberg, D.D.; Reciniello, R.N.

    1998-12-31

    Brookhaven Medical Research Reactor (BMRR) is a 3 mega-watt (MW) heterogeneous, tank-type, light water cooled and moderated, graphite reflected reactor, which was designed for medical and biological studies and became operational in 1959. Over time, the BMRR was modified to provide thermal and epithermal neutron beams suitable for research studies. NCT studies have been performed at both the epithermal neutron irradiation facility (ENIF) on the east side of the BMRR reactor core and the thermal neutron irradiation facility (TNIF) on the west side of the core. Neutron and gamma-ray dosimetry performed from 1994 to the present in both facilities are described and the results are presented and discussed.

  19. MYRRHA a multi-purpose hybrid research reactor for high-tech applications

    SciTech Connect

    Abderrahim, H. A.; Baeten, P.

    2012-07-01

    MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is the flexible experimental accelerator driven system (ADS) in development at SCK-CEN. MYRRHA is able to work both in subcritical (ADS) as in critical mode. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for generation IV (GEN IV) systems, material developments for fusion reactors, radioisotope production and industrial applications, such as Si-doping. MYRRHA will also demonstrate the ADS full concept by coupling the three components (accelerator, spallation target and subcritical reactor) at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow the study of efficient transmutation of high-level nuclear waste. MYRRHA is based on the heavy liquid metal technology and so it will contribute to the development of lead fast reactor (LFR) technology and in critical mode, MYRRHA will play the role of European technology pilot plant in the roadmap for LFR. In this paper the historical evolution of MYRRHA and the rationale behind the design choices is presented and the latest configuration of the reactor core and primary system is described. (authors)

  20. Neutronic performance of the WWR-M research reactor in Ukraine.

    SciTech Connect

    Pond, R. B.; Hanan, N. A.; Matos, J. E.; Mahlers, Y.; Dyakov, A.; Technology Development; Kiev Inst. for Nuclear Research

    2002-01-01

    The 10 MW, WWR-M research reactor of the Kiev Institute for Nuclear Research is jointly studied with the Argonne National Laboratory to examine the feasibility of conversion from HEU (36%) to LEU (19.75%) fuel. A potential core configuration was chosen for comparison of analytical results with HEU fuel and candidate replacement LEU fuels. Core reactivity, fuel assembly power, experiment flux, fuel-cycle length, the number of fuel assemblies consumed per year, and shutdown margins are compared using HEU and LEU fuels. The reactor currently uses HEU (36%) WWR-M2 fuel assemblies (3 tubes, UO2-Al fuel meat with 1.1 gU/cm3 and 37.0 g {sup 235}U). Candidate LEU replacement fuel assemblies, which would result in the same fuel cycle length and the same annual fuel consumption as the HEU (36%) fuel are: LEU WWR-M2 (3 tubes, UO2-Al fuel meat with 2.3 gU/cm3 and 38.3 g 235U) and LEU WWR-MR (37 pins, U9Mo-Al fuel meat with 2.4 gU/cm3 and 38.1 g {sup 235}U). Five LEU WWR-M2 fuel assemblies with 41.7 g {sup 235}U per assembly, UO{sub 2}-Al fuel meat with 2.5 gU/cm3, and a fueled height of 50 cm have completed irradiation testing in the WWR-M reactor at the Petersburg Nuclear Physics Institute in Gatchina to an average 235U burnup of over 70%. This LEU fuel is considered to be qualified for conversion of the WWR-M reactor in Kiev and other research reactors using HEU (36%) WWR-M2 fuel assemblies. For reactors using assemblies with a fueled height of 60 cm, the 235U content per assembly would be 50 g with the same fuel meat composition as the fuel assemblies that were tested in Gatchina. Two 37-pin LEU test assemblies - one with UO{sub 2}-Al fuel meat and about 48 g {sup 235}U and the other with U9Mo-Al fuel meat and about 96 g {sup 235}U are scheduled to begin irradiation testing in the WWR-M reactor in Gatchina before the end of 2002. If these tests (lasting about two years) are successful, LEU pin-type fuel assemblies with up to 96 g 235U would be candidate fuels for LEU

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

    NASA Astrophysics Data System (ADS)

    Sayyah, A.; Rahmani, F.; Khalafi, H.

    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.

  2. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  3. Neutron measurement support for radiation material science in research reactors

    SciTech Connect

    Bregadze, Yu.I.; Grigor'ev, E.I.; Ivanov, V.B.; Klinov, A.V.; Kuprienko, V.A.; Starostov, B.I.; Yaryna, V.P.

    1987-07-01

    The authors discuss the need for development of a support system for unity in neutron measurement. Basic problems are outlined with regard to the support of neutron measurement unity for specialized service, such as research of working neutron fields, monitoring neutron fields during irradiation, and classification, analysis and storage of information about the controlled neutron field. Methods of realizing the metrological functions of the Metrological Center of Neutron Measurements are discussed as is the technical equipment used by the Center.

  4. MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT

    SciTech Connect

    Vinson, D.

    2010-07-11

    The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

  5. 77 FR 26321 - Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-03

    ... Opportunity for Hearing published in the Federal Register on August 19, 2011 (76 FR 52018- 52022). The NRC... Register on March 30, 2012 (77 FR 19362-19366), and concluded that renewal of the facility operating...-water cooled, and shielded TRIGA (Training, Research, Isotope Production, General Atomics) reactor...

  6. Characterization of the fast neutron irradiation facility of the Portuguese Research Reactor after core conversion.

    PubMed

    Marques, J G; Sousa, M; Santos, J P; Fernandes, A C

    2011-08-01

    The fast neutron irradiation facility of the Portuguese Research Reactor was characterized after the reduction in uranium enrichment and rearrangement of the core configuration. In this work we report on the determination of the hardness parameter and the 1MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard. PMID:21071234

  7. Characterization of a Real-time Neutron Imaging Test Station at China Advanced Research Reactor

    NASA Astrophysics Data System (ADS)

    He, Linfeng; Han, Songbai; Wang, Hongli; Wei, Guohai; Wang, Yu; Wu, Meimei; Liu, Yuntao; Chen, Dongfeng

    A real-time neutron imaging test station was recently installed at the China Advanced Research Reactor. The objective of this work was to determine its operational characteristics, including neutron beam profile, the spatial resolution and time resolution. The performance of the equipment was demonstrated by a real time neutron imaging test of the water dynamics in a fuel cell.

  8. Sample Heat, Activity, Reactivity, and Dose Analysis for Safety Analysis of Irradiations in a Research Reactor.

    Energy Science and Technology Software Center (ESTSC)

    1987-12-01

    SHARDA is a program for assessing sample heating rates, activities produced and reactivity load caused while irradiating a small sample in a well thermalized research reactor like CIRUS. It estimates the sample cooling or lead shielding requirements to limit the gamma-ray dose rates due to the irradiated sample within permissible levels.

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

    SciTech Connect

    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.

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

    SciTech Connect

    Not Available

    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.

  11. Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel

    SciTech Connect

    Matos, J.E.; Mo, S.C.; Woodruff, W.L.

    1992-09-01

    This document presents information concerning: analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU; changes to technical specifications mandated by the conversion of the GTRR to low enrichment fuel; changes in the Safety Analysis Report mandated by the conversion of the GTRR to low enrichment fuel; and copies of all changed pages of the SAR and the technical specifications.

  12. Transition phase of the whole-core demonstration at the Oak Ridge Research Reactor

    SciTech Connect

    Hobbs, R.W.; Bretscher, M.M.; Cornella, R.J.; Snelgrove, J.L.

    1986-01-01

    The transition from operation of the Oak Ridge Research Reactor with high-enrichment uranium (HEU) fuel to operation with low-enrichment uranium (LEU) fuel is nearing completion. The systematics of the replacement of the HEU fuel with the LEU fuel are discussed. The results of the core physics measurements that have been conducted during the transition phase are described.

  13. ReactorHealth Physics operations at the NIST center for neutron research.

    PubMed

    Johnston, Thomas P

    2015-02-01

    Performing health physics and radiation safety functions under a special nuclear material license and a research and test reactor license at a major government research and development laboratory encompasses many elements not encountered by industrial, general, or broad scope licenses. This article reviews elements of the health physics and radiation safety program at the NIST Center for Neutron Research, including the early history and discovery of the neutron, applications of neutron research, reactor overview, safety and security of radiation sources and radioactive material, and general health physics procedures. These comprise precautions and control of tritium, training program, neutron beam sample processing, laboratory audits, inventory and leak tests, meter calibration, repair and evaluation, radioactive waste management, and emergency response. In addition, the radiation monitoring systems will be reviewed including confinement building monitoring, ventilation filter radiation monitors, secondary coolant monitors, gaseous fission product monitors, gas monitors, ventilation tritium monitor, and the plant effluent monitor systems. PMID:25551649

  14. Nuclear nonproliferation: Concerns with US delays in accepting foregin research reactors` spent fuel

    SciTech Connect

    1994-03-25

    One key US nonproliferation goal is to discourage use of highly enriched uranium fuel (HEU), which can be used to make nuclear bombs, in civilian nuclear programs worldwide. DOE`s Off-Site Fuels Policy for taking back spent HEU from foreign research reactors was allowed to expire due to environmental reasons. This report provides information on the effects of delays in renewing the Off-Site Fuels Policy on US nonproliferation goals and programs (specifically the reduced enrichment program), DOE`s efforts to renew the fuels policy, and the price to be charged to the operators of foreign reactors for DOE`s activities in taking back spent fuel.

  15. Neutronic safety parameters and transient analyses for potential LEU conversion of the Budapest Research Reactor.

    SciTech Connect

    Pond, R. B.; Hanan, N. A.; Matos, J. E.; Maraczy, C.

    1999-09-27

    An initial safety study for potential LEU conversion of the Budapest Research Reactor was completed. The study compares safety parameters and example transients for reactor cores with HEU and LEU fuels. Reactivity coefficients, kinetic parameters and control rod worths were calculated for cores with HEU(36%) UAl alloy fuel and UO{sub 2}-Al dispersion fuel, and with LEU (19.75%)UO{sub 2}-Al dispersion fuel that has a uranium density of about 2.5 g/cm{sup 3}. A preliminary fuel conversion plan was developed for transition cores that would convert the BRR from HEU to LEU fuel after the process is begun.

  16. Model development experimental programs as part of the NRC reactor safety research

    SciTech Connect

    Young, M.W.; Hsu, Y.Y.

    1982-07-01

    Experimental and model development programs have a key impact on the overall success of code calculational capabilities in addition to supporting regulatory and licensing decisions. The reactor safety research effort undertaken by the Nuclear Regulatory Commission (NRC) has as one of several objectives to obtain experimental data for model and code development and code assessment. This article highlights recent research sponsored under the thermal-hydraulic model-development experimental programs at NRC.

  17. Study of UO/sub 2/ wafer fuel for very high-power research reactors

    SciTech Connect

    Hsieh, T.C.; Jankus, V.Z.; Rest, J.; Billone, M.C.

    1980-11-01

    The Reduced Enrichment Research and Test Reactor Program is aimed at reducing fuel enrichment to < 20% in those research and test reactors presently using highly enriched uranium fuel. UO/sub 2/ caramel fuel is one of the most promising new types of reduced-enrichment fuel for use in research reactors with very high power density. Parametric studies have been carried out to determine the maximum specific power attainable without significant fission-gas release for UO/sub 2/ wafers ranging from 0.75 to 1.50 mm in thickness. The results indicate that (1) all the fuel designs considered in this study are predicted not to fail under full-power operation up to a burnup of 1.09 x 10/sup 21/ fis/cm/sup 3/; (2) for all fuel designs, failure is predicted at approximately the same fuel centerline temperature for a given burnup; (3) the thinner the wafer, and wider the margin for fuel specific power between normal operation and increased-power operation leading to fuel failure; (4) increasing the coolant pressure in the reactor core could improve fuel performance by maintaining the fuel at a higher power level without failure for a given burnup; and (5) for a given power level, fuel failure will occur earlier at a higher cladding surface temperature and/or under power-cycling conditions. 12 figures, 7 tables.

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

    SciTech Connect

    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.

  19. Empirical correlation of residual gamma radiation resulting from operation of the Health Physics Research Reactor

    SciTech Connect

    Chou, T.L.; Ragan, G.E.; Sims, C.S.

    1985-04-01

    An empirical equation has been developed which gives gamma dose equivalent rate as a function of time, distance, and fission yield after a pulsed operation of Oak Ridge National Laboratory's (ORNL) unshielded Health Physics Research Reactor (HPRR). A related expression which is applicable to steady-state reactor operation has been mathematically derived from the aforementioned empirical equation. The two relations can be used to predict the gamma dose equivalent rate to within 25% for times between 1 minute and 90 minutes after reactor shutdown. Similar agreement is expected for up to several days. In most cases the relations are expected to overestimate the gamma dose equivalent rate. 5 refs., 4 figs., 1 tab.

  20. Current status of advanced pelletized cold moderators development for IBR-2M research reactor

    NASA Astrophysics Data System (ADS)

    Kulikov, S.; Belyakov, A.; Bulavin, M.; Mukhin, K.; Shabalin, E.; Verhoglyadov, A.

    2013-03-01

    The world's first advanced pelletized cold neutron moderator is prepared to be put into operation at the IBR-2M pulsed research reactor. It provides long-wavelength neutrons to the most of neutron spectrometers at the beams of the IBR-2M reactor. Aromatic hydrocarbons are used as a material for cold moderators. It is a very attractive material because of its high radiation resistance, good moderating properties, incombustibility, etc. It is shown that the idea of beads transport by a helium flow at cryogenic temperatures is successful. The recent progress and plans for moderator development at the IBR-2M reactor as well as the experimental results of beads transport are discussed in the paper.

  1. The Oak Ridge Research Reactor: safety analysis: Volume 2, supplement 2

    SciTech Connect

    Hurt, S.S.

    1986-11-01

    The Oak Ridge Research Reactor Safety Analysis was last updated via ORNL-4169, Vol. 2, Supplement 1, in May of 1978. Since that date, several changes have been effected through the change-memo system described below. While these changes have involved the cooling system, the electrical system, and the reactor instrumentation and controls, they have not, for the most part, presented new or unreviewed safety questions. However, some of the changes have been based on questions or recommendations stemming from safety reviews or from reactor events at other sites. This paper discusses those changes which were judged to be safety related and which include revisions to the syphon-break system and changes related to seismic considerations which were very recently completed. The maximum hypothetical accident postulated in the original safety analysis requires dynamic containment and filtered flow for compliance with 10CFR100 limits at the site boundary.

  2. The Annular Core Research Reactor (ACRR) postulated limiting event initial and building source terms

    SciTech Connect

    Restrepo, L F

    1992-08-01

    As part of the update of the Safety analysis Report (SAR) for the Annular Core Research Reactor (ACRR), operational limiting events under the category of inadvertent withdrawal of an experiment while at power or during a power pulse were determined to be the most limiting event(s) for this reactor. This report provides a summary of the assumptions, modeling, and results in evaluation of: Reactivity and thermal hydraulics analysis to determine the amount of fuel melt or fuel damage ratios; The reactor inventories following the limiting event; A literature review of post NUREG-0772 release fraction experiment results on severe fuel damages; Decontamination factors due to in-pool transport; and In-building transport modeling and building source term analysis.

  3. Strategic Plan for Light Water Reactor Research and Development

    SciTech Connect

    2004-02-01

    The purpose of this strategic plan is to establish a framework that will allow the Department of Energy (DOE) and the nuclear power industry to jointly plan the nuclear energy research and development (R&D) agenda important to achieving the Nation's energy goals. This strategic plan has been developed to focus on only those R&D areas that will benefit from a coordinated government/industry effort. Specifically, this plan focuses on safely sustaining and expanding the electricity output from currently operating nuclear power plants and expanding nuclear capacity through the deployment of new plants. By focusing on R&D that addresses the needs of both current and future nuclear plants, DOE and industry will be able to take advantage of the synergism between these two technology areas, thus improving coordination, enhancing efficiency, and further leveraging public and private sector resources. By working together under the framework of this strategic plan, DOE and the nuclear industry reinforce their joint commitment to the future use of nuclear power and the National Energy Policy's goal of expanding its use in the United States. The undersigned believe that a public-private partnership approach is the most efficient and effective way to develop and transfer new technologies to the marketplace to achieve this goal. This Strategic Plan is intended to be a living document that will be updated annually.

  4. Development and operation of research-scale III-V nanowire growth reactors

    NASA Astrophysics Data System (ADS)

    Schroer, M. D.; Xu, S. Y.; Bergman, A. M.; Petta, J. R.

    2010-02-01

    III-V nanowires are useful platforms for studying the electronic and mechanical properties of materials at the nanometer scale. However, the costs associated with commercial nanowire growth reactors are prohibitive for most research groups. We developed hot-wall and cold-wall metal organic vapor phase epitaxy reactors for the growth of InAs nanowires, which both use the same gas handling system. The hot-wall reactor is based on an inexpensive quartz tube furnace and yields InAs nanowires for a narrow range of operating conditions. Improvement of crystal quality and an increase in growth run to growth run reproducibility are obtained using a homebuilt UHV cold-wall reactor with a base pressure of 2×10-9 Torr. A load lock on the UHV reactor prevents the growth chamber from being exposed to atmospheric conditions during sample transfers. Nanowires grown in the cold-wall system have a low defect density, as determined using transmission electron microscopy, and exhibit field effect gating with mobilities approaching 16 000 cm2/(V s).

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

    NASA Astrophysics Data System (ADS)

    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.

  6. The dependence of the risk of research reactors from their operating schedule--a case study.

    PubMed

    Kollas, J G

    1991-06-01

    The dependence of the radiological consequences of the "Demokritos" research reactor on the operating schedule of the reactor is assessed in this paper. The 5 MW reactor is located within the limits of Athens city, a large population center with over 3 million inhabitants. The consequences examined would be due to the occurrence of a postulated accident, a 20% core melt loss of coolant accident, that is also considered as the design basis accident of the reactor. Three operating schedules are taken into account: (a) a continuous operation schedule; (b) a 16 hr/day, 5 days/week schedule; and (c) the present 8 hr/day, 5 days/week operating schedule. The assessment of the source term emerges from a conservative estimation of fission product releases to the reactor operating floor, and further under the conservative assumptions of no filter mitigation, and a ground release to the environment. The results of the analysis indicate that there is a direct relation between consequences and duration of operation, the former becoming appreciable as the continuous operation limit is approached. In all cases examined, the thyroid dose and the latent thyroid health effects would be the limiting consequences. PMID:1876720

  7. Fresh and Spent Nuclear Fuel Repatriation from the IRT-2000 Research Reactor Facility, Sofia, Bulgaria

    SciTech Connect

    K. J. Allen; T. G. Apostolov; I. S. Dimitrov

    2009-03-01

    The IRT 2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned.

  8. Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

    SciTech Connect

    Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh; Ba Vien Luong; Kien Cuong Nguyen

    2008-07-15

    The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configuration with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)

  9. Development of Regulatory Technical Requirements for the Advanced Integral Type Research Reactor

    SciTech Connect

    Jo, Jong Chull; Yune, Young Gill; Kim, Woong Sik; Kim, Hho Jung

    2004-07-01

    This paper presents the current status of the study on the development of regulatory technical requirements for the licensing review of an advanced integral type research reactor of which the license application is expected in a few years. According to the Atomic Energy Act of Korea, both research and education reactors are subject to the technical requirements for power reactors in the licensing review. But, some of the requirements may not be applicable or insufficient for the licensing reviews of reactors with unique design features. Thus it is necessary to identify which review topics or areas can not be addressed by the existing requirements and to develop the required ones newly or supplement appropriately. Through the study performed so far, it has been identified that the following requirements need to be developed newly for the licensing review of SMART-P: the use of proven technology, the interfacial facility, the non-safety systems, and the metallic fuels. The approach and basis for the development of each of the requirements are discussed. (authors)

  10. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    SciTech Connect

    Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2013-07-01

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  11. Corrosion Surveillance for Research Reactor Spent Nuclear Fuel in Wet Basin Storage

    SciTech Connect

    Howell, J.P.

    1998-10-16

    Foreign and domestic test and research reactor fuel is currently being shipped from locations over the world for storage in water filled basins at the Savannah River Site (SRS). The fuel was provided to many of the foreign countries as a part of the "Atoms for Peace" program in the early 1950's. In support of the wet storage of this fuel at the research reactor sites and at SRS, corrosion surveillance programs have been initiated. The International Atomic Energy Agency (IAEA) established a Coordinated Research Program (CRP) in 1996 on "Corrosion of Research Reactor Aluminum-Clad Spent Fuel in Water" and scientists from ten countries worldwide were invited to participate. This paper presents a detailed discussion of the IAEA sponsored CRP and provides the updated results from corrosion surveillance activities at SRS. In May 1998, a number of news articles around the world reported stories that microbiologically influenced corrosion (MIC) was active on the aluminum-clad spent fuel stored in the RBOF basin at SRS. This assessment was found to be in error with details presented in this paper. A biofilm was found on aluminum coupons, but resulted in no corrosion. Cracks seen on the surface were not caused by corrosion, but by stresses from the volume expansion of the oxide formed during pre-conditioning autoclaving. There has been no pitting caused by MIC or any other corrosion mechanism seen in the RBOF basin since initiation of the SRS Corrosion Surveillance Program in 1993.

  12. Corrosion surveillance for research reactor spent nuclear fuel in wet basin storage

    SciTech Connect

    Howell, J.P.

    1999-07-01

    Foreign and domestic test and research reactor fuel is currently being shipped from locations over the world for storage in water filled basins at the Savannah River Site (SRS). The fuel was provided to many of the foreign countries as a part of the ``Atoms for Peace'' program in the early 1950's. In support of the wet storage of this fuel at the research reactor sites and at SRS, corrosion surveillance programs have been initiated. The International Atomic Energy Agency (IAEA) established a Coordinated Research Program (CRP) in 1996 on ``Corrosion of Research Reactor Aluminum-Clad Spent Fuel in Water'' and scientists from ten countries worldwide were invited to participate. This paper presents a detailed discussion of the IAEA sponsored CRP and provides the updated results from corrosion surveillance activities at SRS. In May 1998, a number of news articles around the world reported stories that microbiologically influenced corrosion (MIC) was active on the aluminum-clad spent fuel stored in the Receiving Basin for Offsite Fuels (RBOF) at SRS. This assessment was found to be in error with details presented in this paper. A biofilm was found on aluminum coupons, but resulted in no corrosion. Cracks seen on the surface were not caused by corrosion, but by stresses from the volume expansion of the oxide formed during pre-conditioning autoclaving. There has been no pitting caused by MIC or any other corrosion mechanism seen in the RBOF basin since initiation of the SRS Corrosion Surveillance Program in 1993.

  13. U.S. Department of Energy Instrumentation and Controls Technology Research for Advanced Small Modular Reactors

    SciTech Connect

    Wood, Richard Thomas

    2012-01-01

    Instrumentation, controls, and human-machine interfaces (ICHMI) are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The U.S. Department of Energy (DOE) has recognized that ICHMI research, development, and demonstration (RD&D) is needed to resolve the technical challenges that may compromise the effective and efficient utilization of modern ICHMI technology and consequently inhibit realization of the benefits offered by expanded utilization of nuclear power. Consequently, key DOE programs have substantial ICHMI RD&D elements to their respective research portfolio. This article describes current ICHMI research to support the development of advanced small modular reactors.

  14. A human reliability analysis of the University of New Mexico`s AGN- 201M nuclear research reactor. Revision 1

    SciTech Connect

    Brumburgh, G.P.; Heger, A.S.

    1992-10-15

    During 1990--1991, a probabilistic risk assessment was conducted on the University of New Mexico`s AGN-201M nuclear research reactor to address the risk and consequence of a maximum hypothetical release accident. The assessment indicated a potential for consequential human error to precipitate Chis scenario. Subsequently, a human reliability analysis was performed to evaluate the significance of human interaction on the reactor`s safety systems. This paper presents the results of that investigation.

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

    PubMed

    Lashkari, A; Khalafi, H; Kazeminejad, H

    2013-05-01

    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

  16. The key-role of instrumentation for the new generation of research reactors

    SciTech Connect

    Bignan, G.; Villard, J. F.; Destouches, C.; Baeten, P.; Vermeeren, L.; Michiels, S.

    2011-07-01

    Experimental reactors have been indispensable since the beginning of the use of nuclear energy to support many important fields of industry and research: safety, lifetime management and operation optimisation of nuclear power plants, development of new types of reactors with improved resources and fuel cycle management, medical applications, material development for fusion... Over the last decade, modifications of the operational needs and the ageing of the nuclear facilities have led to several closures and time is coming for new key European Experimental Reactors (EER) within a European and International Framework. Projects like MYRRHA and JHR are underway to define and implement a new consistent EER policy: - Meeting industry and public needs, keeping a high level of scientific expertise; - With a limited number of EER, specified within a rational compromise between specialisation, complementarities and back-up capacities; - To be put into effective operation in this or the next decade. These new projects will give to the scientific community high performances allowing innovative fields of R and D. A new generation of instrumentation to address new phenomena and that allows better on-line investigation of some key physical parameters is necessary to achieve these challenges. One initiative to progress in this direction is the Joint Instrumentation Laboratory between CEA and SCK.CEN which has already given significant results and patents. Major scientific challenges to achieve in the field of instrumentation for this new generation of European Research Reactors have to be investigated and are described in this paper as well as a short description of the JHR and MYRRHA reactors that will be serving as flexible irradiation facilities for testing them. (authors)

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

    SciTech Connect

    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.

  18. Decommissioning of German Research Reactors Under the Governance of the Federal Ministry of Education and Research - 12154

    SciTech Connect

    Weigl, M.

    2012-07-01

    Since 1956, nuclear research and development (R and D) in Germany has been supported by the Federal Government. The goal was to help German industry to become competitive in all fields of nuclear technology. National research centers were established and demonstration plants were built. In the meantime, all these facilities were shut down and are now in a state of decommissioning and dismantling (D and D). Meanwhile, Germany is one of the leading countries in the world in the field of D and D. Two big demonstration plants, the Niederaichbach Nuclear Power Plant (KKN) a heavy-water cooled pressure tube reactor with carbon-dioxide cooling and the Karlstein Superheated Steam Reactor (HDR) a boiling light water reactor with a thermal power of 100 MW, are totally dismantled and 'green field' is reached. Another big project was finished in 2008. The Forschungs-Reaktor Juelich 1 (FRJ1), a research reactor with a thermal power of 10 MW was completely dismantled and in September 2008 an oak tree was planted on a green field at the site, where the FRJ1 was standing before. This is another example for German success in the field of D and D. Within these projects a lot of new solutions and innovative techniques were tested, which were developed at German universities and in small and medium sized companies mostly funded by the Federal Ministry of Education and Research (BMBF). Some examples are underwater-cutting technologies like plasma arc cutting and contact arc metal cutting. This clearly shows that research on the field of D and D is important for the future. Moreover, these research activities are important to save the know-how in nuclear engineering in Germany and will enable enterprises to compete on the increasing market of D and D services. The author assumes that an efficient decommissioning of nuclear installations will help stabilize the credibility of nuclear energy. Some critics of nuclear energy are insisting that a return to 'green field sites' is not possible

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

    SciTech Connect

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

    SciTech Connect

    Yamada, K.; Aksan, S. N.

    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)

  1. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    SciTech Connect

    Pond, R.B.; Matos, J.E.

    1996-12-31

    This document has been prepared to assist research reactor operators possessing spent fuel containing enriched uranium of United States origin to prepare part of the documentation necessary to ship this fuel to the United States. Data are included on the nuclear mass inventory, photon dose rate, and thermal decay heat of spent research reactor fuel assemblies. Isotopic masses of U, Np, Pu and Am that are present in spent research reactor fuel are estimated for MTR, TRIGA and DIDO-type fuel assembly types. The isotopic masses of each fuel assembly type are given as functions of U-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly. Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg[sup 235]U, and for fission product decay times of up to 20 years. Thermal decay heat loads are estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time.

  2. Progress made on the University of Missouri research reactor HEU to LEU fuel conversion feasibility study

    SciTech Connect

    McKibben, J. Charles; Kutikkad, Kiratadas; Foyto, Leslie P.

    2008-07-15

    The University of Missouri Research Reactor (MURR), the highest-powered University-owned research reactor in the U.S. designed to operate at a maximum steady-state power level of 10 MW{sub th}, is one of five U.S. high performance research reactors that use HEU fuel that is actively th collaborating with the U.S. Department of Energy to find a suitable LEU fuel replacement. A conversion feasibility study, using U-10Mo monolithic LEU fuel, is currently being performed at MURR. At first, broad scoping studies where conducted using the transport code MCNP, where the core water-to-metal ratio was varied by altering the thickness or width of the plate cladding, fuel meat, and coolant channel gaps, and varying the number of fuel plates. From these studies, an optimal LEU core design was chosen based on the following calculated parameters: power peaking factors, excess reactivity, and the fast and thermal fluxes available to the experimental facilities. Fuel burnup calculations are now being performed using the 3-D diffusion theory code REBUS. Also included in this paper are some preliminary safety analyses, including parametric studies using the reactivity transient code PARET-ANL and hydraulic calculations using the light- and heavy-water thermal-hydraulic transient code RELAP5/MOD3.3. (author)

  3. The RERTR (Reduced Enrichment Research and Test Reactor) program: A progress report

    SciTech Connect

    Travelli, A.

    1986-11-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1985, the activities, results, and new developments which occurred in 1986 are reviewed. The second miniplate series, concentrating on U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al fuels, was expanded and its irradiation continued. Postirradiation examinations of several of these miniplates and of six previously irradiated U/sub 3/Si/sub 2/-Al full-size elements were completed with excellent results. The whole-core ORR demonstration with U/sub 3/Si/sub 2/-Al fuel at 4.8 g U/cm/sup 3/ is well under way and due for completion before the end of 1987. DOE removed an important barrier to conversions by announcing that the new LEU fuels will be accepted for reprocessing. New DOE prices for enrichment and reprocessing services were calculated to have minimal effect on HEU reactors, and to reduce by about 8 to 10% the total fuel cycle costs of LEU reactors. New program activities include preliminary feasibility studies of LEU use in DOE reactors, evaluation of the feasibility to use LEU targets for the production of fission-product /sup 99/Mo, and responsibility for coordinating safety evaluations related to LEU conversions of US university reactors, as required by NRC. Achievement of the final program goals is projected for 1990. This progress could not have been achieved without close international cooperation, whose continuation and intensification are essential to the achievement of the ultimate goals of the RERTR Program.

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

    SciTech Connect

    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.

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

    SciTech Connect

    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.

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

    SciTech Connect

    West, C.D.

    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.

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

    NASA Technical Reports Server (NTRS)

    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.

  8. Design of a Thermal Neutron Beam for a New Neutron Imaging Facility at Tehran Research Reactor

    NASA Astrophysics Data System (ADS)

    Dastjerdi, Mohammad Hossein Choopan; Khalafi, Hossein

    A new neutron imaging facility will be built around the Tehran Research Reactor (TRR). The TRR is an open pool light water moderated5 MW research reactor with six beam tubes. The neutron energy spectrum near the reactor core at the entrance of the beam tube was measured by the foil activation method using the SAND-II code and calculated by the MCNP Monte Carlo code. There was a good similarity between calculated and simulated spectra. The principal component of this facility is its neutron collimator. The collimator is a beam-forming assembly which determines the geometric properties of the beam. In addition, it may contain filters to modify the energy spectrum or to reduce the gamma ray content of the beam. The optimum thickness of filters, the position of the aperture and other details of the neutron collimator were calculated using MCNP Monte Carlo simulations. In this design, the L/D ratio of this facility had the value of 120. The thermal neutron flux at the image plane was about 7.8×106 n/cm2.s and n/γ ratio about 106 n/cm2.μSv.

  9. The Advanced Neutron Source (ANS) project: A world-class research reactor facility

    SciTech Connect

    Thompson, P.B.; Meek, W.E.

    1993-07-01

    This paper provides an overview of the Advanced Neutron Source (ANS), a new research facility being designed at Oak Ridge National Laboratory. The facility is based on a 330 MW, heavy-water cooled and reflected reactor as the neutron source, with a thermal neutron flux of about 7.5{times}10{sup 19}m{sup {minus}2}{center_dot}sec{sup {minus}1}. Within the reflector region will be one hot source which will serve 2 hot neutron beam tubes, two cryogenic cold sources serving fourteen cold neutron beam tubes, two very cold beam tubes, and seven thermal neutron beam tubes. In addition there will be ten positions for materials irradiation experiments, five of them instrumented. The paper touches on the project status, safety concerns, cost estimates and scheduling, a description of the site, the reactor, and the arrangements of the facilities.

  10. Decommissioning Small Research and Training Reactors; Experience on Three Recent University Projects - 12455

    SciTech Connect

    Gilmore, Thomas; DeWitt, Corey; Miller, Dustin; Colborn, Kurt

    2012-07-01

    Decommissioning small reactors within the confines of an active University environment presents unique challenges. These range from the radiological protection of the nearby University population and grounds, to the logistical challenges of working in limited space without benefit of the established controlled, protected, and vital areas common to commercial facilities. These challenges, and others, are discussed in brief project histories of three recent (calendar year 2011) decommissioning activities at three University training and research reactors. These facilities include three separate Universities in three states. The work at each of the facilities addresses multiple phases of the decommissioning process, from initial characterization and pre-decommissioning waste removal, to core component removal and safe storage, through to complete structural dismantlement and site release. The results of the efforts at each University are presented, along with the challenges that were either anticipated or discovered during the decommissioning efforts, and results and lessons learned from each of the projects. (authors)

  11. Design of the cold neutron triple-axis spectrometer at the China Advanced Research Reactor

    NASA Astrophysics Data System (ADS)

    Cheng, P.; Zhang, Hongxia; Bao, W.; Schneidewind, A.; Link, P.; Grünwald, A. T. D.; Georgii, R.; Hao, L. J.; Liu, Y. T.

    2016-06-01

    The design of the first cold neutron triple-axis spectrometer at the China Advanced Research Reactor is presented. Based on the Monte Carlo simulations using neutron ray-tracing program McStas, the parameters of major neutron optics in this instrument are optimized. The neutron flux at sample position is estimated to be 5.6 ×107 n/cm2/s at neutron incident energy Ei=5 meV when the reactor operates normally at the designed 60 MW power. The performances of several neutron supermirror polarizing devices are compared and their critical parameters are optimized for this spectrometer. The polarization analysis will be realized with a flexible switch from the unpolarized experimental mode.

  12. Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

    PubMed

    Mohammadi, A; Hassanzadeh, M; Gharib, M

    2016-02-01

    In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified. PMID:26720262

  13. Flux measurements in a nuclear research reactor by using an aluminum nitride detector

    NASA Astrophysics Data System (ADS)

    Moon, B. S.; Yoo, D. S.; Hwang, I. K.; Chung, C. E.; Holcomb, D. E.

    2007-08-01

    A small polycrystalline aluminium nitride detector with a thickness of 381 μm was used to measure a 200,000 Ci Co 60 source and to measure the flux in a research reactor where the neutron flux is about 10 14/cm 2 s, which is nearly the same order as in the commercial power plant. If the applied voltage is greater than or equal to 2000 V and if the measurements are done in a short period of time so that the heat energy does not build up in the aluminium nitride, then the measured electric current is linearly proportional to the input flux. It is assumed of course that the energy spectrum of the input flux remains constant. This linearity relation is illustrated by the results of a measurement in which the reactor power has been controlled so that the flux becomes a step function.

  14. Microstructure of 50 year old SCK CEN BR1 research reactor fuel

    SciTech Connect

    Leenaers, A.; Berghe, S. van den

    2008-07-15

    The BR1 research reactor at SCK CEN, Mol (Belgium) has a graphite core matrix loaded with fuel rods consisting of a natural uranium slug in an aluminum cladding. Fabrication reports show the application of a so-called AlSi bonding layer and an U(Al,Si){sub 3} anti-diffusion layer on the natural uranium fuel slug to limit the interaction between the uranium fuel and aluminum cladding. The BR1 reactor is in operation since 1956 and still contains its original fuel rods. After more than 50 years irradiation at low temperature, the integrity of some of the fuel rods is investigated. The microstructure of the fuel, bonding and anti-diffusion layer and cladding is analysed using optical microscopy (OM), scanning electron microscopy (SEM) and electron microprobe analysis (EPMA). (author)

  15. A SF6 RF plasma reactor for research on textile treatment

    NASA Astrophysics Data System (ADS)

    Riccardi, C.; Barni, R.; Fontanesi, M.; Marcandalli, B.; Massafra, M.; Selli, E.; Mazzone, G.

    2001-02-01

    This research concerns the development of a SF6 RF discharge at low pressure in a small reactor for industrial applications. The plasma is produced in the pressure range 0.05-1 mbar by a RF supply. The pumping system sustains a flowrate of about 50 cm s-1, with residence time in the discharge of about 0.2 s at a pressure of 0.1 mbar. The discharge parameters were measured at a low operation power. Measurements were performed by means of movable electrostatic probes and a photodiode. Particular care in the analysis of the data proved to be necessary due to the presence of a substantial amount of negative ions. The reactor has been employed for textile treatment in order to modify the surface properties of the fibres. Favourable operating conditions leading to an improved hydrophobicity of the textiles were achieved.

  16. RELAP5 Application to Accident Analysis of the NIST Research Reactor

    SciTech Connect

    Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.; Diamond, D.

    2012-03-18

    Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accident and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.

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

    SciTech Connect

    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

  18. Regulatory Experiences for the Decommissioning of the Research Reactor in Korea

    SciTech Connect

    CHOI, Kyung-Woo

    2008-01-15

    The first research reactor in Korea (KRR-1, TRIGA Mark-II) has operated since 1962, and the second one (KRR-2, TRIGA Mark-III), since 1972. Both of them were phased out in 1995 due to their lives and the operation of a new research reactor, HANARO (30 MW thermal power) operated by KAERI (Korea Atomic Energy Research Institute). After deciding the shutdown by the Nuclear Development and Utilization Committee in March 1996, KAERI began to prepare the decommissioning plan, including the environmental impact assessment, and submitted the plan to the Ministry of Science and Technology (MOST) in December 1998. Korea Institute of Nuclear Safety (KINS) reviewed document and prepared the review report in 1999. KINS is an organization of technical expertise which performs regulatory functions, entrusted by the MOST in accordance with the Atomic Energy Act and its Enforcement Decree. The review report written by KINS was consulted by the Special Committee on Nuclear Safety in January 2000. The committee submitted their consultation results to the Nuclear Safety Commission for the final approval by the Minister of MOST. The license was issued in November 2000. With the consent of the Korean government to the US Record of Decision, the spent fuel of KRR-1 and 2 was safely transported to the United States in July 1998. The decontamination and dismantling of KRR-2 was completed at the end of 2005 but the decommissioning of KRR-1 has been suspended by the problem for the memorial of the reactor. After the decommissioning of the research reactor is finished, the site will be returned to the site owner, Korea Electric Power Corporation (KEPCO). In this paper, the state-of-art and lessons learnt from recent regulatory activities for decommissioning of KRR- 2 are summarized. In conclusion: since the shutdown of KRR-1 and 2 had been decided, the safe assessment and licensing review were carried out after applying for decommissioning plan of those research reactors by operator. Through

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

    PubMed

    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. PMID:20855215

  20. International Atomic Energy Agency support of research reactor highly enriched uranium to low enriched uranium fuel conversion projects

    SciTech Connect

    Bradley, E.; Adelfang, P.; Goldman, I.N.

    2008-07-15

    The IAEA has been involved for more than twenty years in supporting international nuclear non- proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly assisted efforts to convert research reactors from HEU to LEU fuel. HEU to LEU fuel conversion projects differ significantly depending on several factors including the design of the reactor and fuel, technical needs of the member state, local nuclear infrastructure, and available resources. To support such diverse endeavours, the IAEA tailors each project to address the relevant constraints. This paper presents the different approaches taken by the IAEA to address the diverse challenges involved in research reactor HEU to LEU fuel conversion projects. Examples of conversion related projects in different Member States are fully detailed. (author)

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

    SciTech Connect

    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.

  2. Production and study of radionuclides at the research institute of atomic reactors (NIIAR)

    SciTech Connect

    Karelin, E.A.; Gordeev, Y.N.; Filimonov, V.T.

    1995-01-01

    The main works of the Radionuclide Sources and Preparations Department (ORIP) of the Research Institute of Atomic Reactors (NIIAR) are summarized. The major activity of the Radionuclide Sources and Preparations Department (ORIP) is aimed at production of radioactive preparations of trans-plutonium elements (TPE) and also of lighter elements (from P to Ir), manufacture of ionizing radiation sources thereof, and scientific research to develop new technologies. One of the radionuclides that recently has received major attention is gadolinium-153. Photon sources based on it are used in densimeters for diagnostics of bone deseases. The procedure for separating gadolinium and europium, which is currently used at the Research Institute of Atomic Reactors (NILAR), is based on europium cementation with the use of sodium amalgam. The method, though efficient, did not until recently permit an exhaustive removal of radioactive europium from {sup 153}Gd. The authors have thoroughly studied the separation process in semi-countercurrent mode, using citrate solutions. A special attention was given to the composition of europium complex species.

  3. Progress of the RERTR (Reduced Enrichment Research and Test Reactor) Program in 1989

    SciTech Connect

    Travelli, A.

    1989-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1988, the major events, findings, and activities of 1989 are reviewed. The scope of the RERTR Program activities was curtailed, in 1989, by an unexpected legislative restriction which limited the ability of the Arms Control and Disarmament Agency to adequately fund the program. Nevertheless, the thrust of the major planned program activities was maintained, and meaningful results were obtained in several areas of great significance for future work. 15 refs., 12 figs.

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

    SciTech Connect

    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. Filtered epithermal quasi-monoenergetic neutron beams at research reactor facilities.

    PubMed

    Mansy, M S; Bashter, I I; El-Mesiry, M S; Habib, N; Adib, M

    2015-03-01

    Filtered neutron techniques were applied to produce quasi-monoenergetic neutron beams in the energy range of 1.5-133keV at research reactors. A simulation study was performed to characterize the filter components and transmitted beam lines. The filtered beams were characterized in terms of the optimal thickness of the main and additive components. The filtered neutron beams had high purity and intensity, with low contamination from the accompanying thermal emission, fast neutrons and γ-rays. A computer code named "QMNB" was developed in the "MATLAB" programming language to perform the required calculations. PMID:25544666

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

    PubMed

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

    2015-05-01

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

  7. Monochromatic Neutron Tomography Using 1-D PSD Detector at Low Flux Research Reactor

    NASA Astrophysics Data System (ADS)

    Ashari, N. Abidin; Saleh, J. Mohamad; Abdullah, M. Zaid; Mohamed, A. Aziz; Azman, A.; Jamro, R.

    2008-03-01

    This paper describes the monochromatic neutron tomography experiment using the 1-D Position Sensitive Neutron Detector (PSD) located at Nuclear Malaysia TRIGA MARK II Research reactor. Experimental work was performed using monochromatic neutron source from beryllium filter and HOPG crystal monochromator. The principal main aim of this experiment was to test the detector efficiency, image reconstruction algorithm and the usage of 0.5 nm monochromatic neutrons for the neutron tomography setup. Other objective includes gathering important parameters and features to characterize the system.

  8. Feasibility study for production of I-131 radioisotope using MNSR research reactor.

    PubMed

    Elom Achoribo, A S; Akaho, Edward H K; Nyarko, Benjamin J B; Osae Shiloh, K D; Odame Duodu, Godfred; Gibrilla, Abass

    2012-01-01

    A feasibility study for (131)I production using a Low Power Research Reactor was conducted to predict the yield of (131)I by cyclic activation technique. A maximum activity of 5.1GBq was achieved through simulation using FORTRAN 90, for an irradiation of 6h. But experimentally only 4h irradiation could be done, which resulted in an activity of 4.0×10(5)Bq. The discrepancy in the activities was due to the fact that beta decays released during the process could not be considered. PMID:21900016

  9. Monochromatic Neutron Tomography Using 1-D PSD Detector at Low Flux Research Reactor

    SciTech Connect

    Ashari, N. Abidin; Saleh, J. Mohamad; Abdullah, M. Zaid; Mohamed, A. Aziz; Azman, A.; Jamro, R.

    2008-03-17

    This paper describes the monochromatic neutron tomography experiment using the 1-D Position Sensitive Neutron Detector (PSD) located at Nuclear Malaysia TRIGA MARK II Research reactor. Experimental work was performed using monochromatic neutron source from beryllium filter and HOPG crystal monochromator. The principal main aim of this experiment was to test the detector efficiency, image reconstruction algorithm and the usage of 0.5 nm monochromatic neutrons for the neutron tomography setup. Other objective includes gathering important parameters and features to characterize the system.

  10. Evaluation of differential shim rod worth measurements in the Oak Ridge Research Reactor

    SciTech Connect

    Bretscher, M.M.

    1987-01-01

    Reasonable agreement between calculated and measured differential shim rod worths in the Oak Ridge Research Reactor (ORR) has been achieved by taking into account the combined effects of negative reactivity contributions from changing fuel-moderator temperatures and of delayed photoneutrons. A method has been developed for extracting the asymptotic period from the shape of the initial portion of the measured time-dependent neutron flux profile following a positive reactivity insertion. In this region of the curve temperature-related reactivity feedback effects are negligibly small. Results obtained by applying this technique to differential shim rod worth measurements made in a wide variety of ORR cores are presented.