WorldWideScience

Sample records for research reactor performance

  1. Performance of a multipurpose research electrochemical reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  2. A new high performance research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abbate, Pablo M. [INVAP S.E., Bariloche, Rio Negro (Argentina)

    2002-07-01

    A contract for the design, construction and commissioning of the Replacement Research Reactor was signed in July 2000 between Australia authorities and INVAP from Argentina. Since then the detailed design has been completed, an application for a construction license was made in May 2001 and the construction authorisation was issued on 4{sup th} April 2002. This paper explains the safety philosophy embedded into the design together with the approach taken for main elements of the design and their relation to the proposed applications of the reactor. Also information is provided on the suit of neutron beam facilities and irradiation facilities being constructed. Finally it is presented an outline of the project management organisation, project planing and schedule. (author)

  3. A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

  4. Cooling Performance of Natural Circulation for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

  5. Study for improvement of performance of the test and research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Fumio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    Current utilization needs for the test and research reactors become more advanced and diversified along with the advance of nuclear science and technology. Besides, the requested safety for the research and test reactors grows strictly every year as well as a case of the power reactors. Under this circumstance, every effort to improve reactor performance including its safety is necessary to be sustained for allowing more effective utilization of the test and research reactors as experimental apparatus for advanced researches. In this study, the following three themes i.e., JMTR high-performance fuel element, evaluation method of fast neutron irradiation dose in the JMTR, evaluation method of performance of siphon break valve as core covering system for water-cooled test and research reactors, were investigated respectively from the views of improvement of core performance as a neutron source, utilization performance as an experimental apparatus, and safety as a reactor plant. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-03-01

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

  7. Multi purpose research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Raina, V.K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)]. E-mail: vkrain@magnum.barc.ernet.in; Sasidharan, K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sengupta, Samiran [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Singh, Tej [Research Reactor Services Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2006-04-15

    At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor.

  8. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

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

  9. INVAP's Research Reactor Designs

    Directory of Open Access Journals (Sweden)

    Eduardo Villarino

    2011-01-01

    Full Text Available INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors.

  10. Aqueous processing of U-10Mo scrap for high performance research reactor fuel

    Science.gov (United States)

    Youker, Amanda J.; Stepinski, Dominique C.; Maggos, Laura E.; Bakel, Allen J.; Vandegrift, George F.

    2012-08-01

    The Global Threat Reduction Initiative (GTRI) Conversion program, which is part of the US government's National Nuclear Security Administration (NNSA), supports the conversion of civilian use of highly enriched uranium (HEU) to low enriched uranium (LEU) for reactor fuel and targets. The reason for conversion is to eliminate the use of any material that may pose a threat to the United States or other foreign countries. High performance research reactors (HPRRs) cannot make the conversion to a standard LEU fuel because they require a more dense fuel to meet their performance requirements. As a result, a more dense fuel consisting of a monolithic uranium-molybdenum alloy containing 10% (w/w) Mo with Al cladding and a Zr bonding-layer is being considered. Significant losses are expected in the fabrication of this fuel, so a means to recycle the scrap pieces is needed. Argonne National Laboratory has developed an aqueous-processing flowsheet for scrap recovery in the fuel fabrication process for high-density LEU-monolithic fuel based on data found in the literature. Experiments have been performed to investigate dissolution conditions for solutions containing approximately 20 g-U/L and 50 g-U/L with and without Fe(NO3)3. HNO3 and HF concentrations have been optimized for timely dissolution of the fuel scrap and prevention of the formation of the U-Zr2 intermetallic, explosive complex, while meeting the requirements needed for further processing.

  11. A study on improving the performance of a research reactor's equilibrium core

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2013-01-01

    Full Text Available Utilizing low enriched uranium silicide fuel (U3Si2-Al of existing uranium density (3.285 g/cm3, different core configurations have been studied in search of an equilibrium core with an improved performance for the Pakistan Research Reactor-1. Furthermore, we have extended our analysis to the performance of higher density silicide fuels with a uranium density of 4.0 and 4.8 U g/cm3. The criterion used in selecting the best performing core was that of “unit flux time cycle length per 235U mass per cycle”. In order to analyze core performance by improving neutron moderation, utilizing higher-density fuel, the effect of the coolant channel width was also studied by reducing the number of plates in the standard/control fuel element. Calculations employing computer codes WIMSD/4 and CITATION were performed. A ten energy group structure for fission neutrons was used for the generation of microscopic cross-sections through WIMSD/4. To search the equilibrium core, two-dimensional core modelling was performed in CITATION. Performance indicators have shown that the higher-density uranium silicide-fuelled core (U density 4.8 g/cm3 without any changes in standard/control fuel elements, comprising of 15 standard and 4 control fuel elements, is the best performing of all analyzed cores.

  12. Aqueous processing of U-10Mo scrap for high performance research reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Youker, Amanda J., E-mail: youker@anl.gov [Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Stepinski, Dominique C.; Maggos, Laura E.; Bakel, Allen J.; Vandegrift, George F. [Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer GTRI program supports conversion from HEU to LEU. Black-Right-Pointing-Pointer High performance research reactors require a dense LEU fuel such as U-10Mo foils. Black-Right-Pointing-Pointer Dissolution conditions for U-10Mo foils in acidic media have been optimized. Black-Right-Pointing-Pointer Solvent-extraction processing can be used to recover U lost in fuel fabrication. Black-Right-Pointing-Pointer Flowsheets were developed using Argonne-design contactors but other contactors can be used as well. - Abstract: The Global Threat Reduction Initiative (GTRI) Conversion program, which is part of the US government's National Nuclear Security Administration (NNSA), supports the conversion of civilian use of highly enriched uranium (HEU) to low enriched uranium (LEU) for reactor fuel and targets. The reason for conversion is to eliminate the use of any material that may pose a threat to the United States or other foreign countries. High performance research reactors (HPRRs) cannot make the conversion to a standard LEU fuel because they require a more dense fuel to meet their performance requirements. As a result, a more dense fuel consisting of a monolithic uranium-molybdenum alloy containing 10% (w/w) Mo with Al cladding and a Zr bonding-layer is being considered. Significant losses are expected in the fabrication of this fuel, so a means to recycle the scrap pieces is needed. Argonne National Laboratory has developed an aqueous-processing flowsheet for scrap recovery in the fuel fabrication process for high-density LEU-monolithic fuel based on data found in the literature. Experiments have been performed to investigate dissolution conditions for solutions containing approximately 20 g-U/L and 50 g-U/L with and without Fe(NO{sub 3}){sub 3}. HNO{sub 3} and HF concentrations have been optimized for timely dissolution of the fuel scrap and prevention of the formation of the U-Zr{sub 2} intermetallic, explosive complex, while

  13. INVAP's Research Reactor Designs

    OpenAIRE

    Eduardo Villarino; Alicia Doval

    2011-01-01

    INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper ...

  14. Performance improvement of artificial neural networks designed for safety key parameters prediction in nuclear research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mazrou, Hakim [Division de Physique Radiologique, Centre de Recherche Nucleaire d' Alger (CRNA), 02 Boulevard Frantz, Fanon, B.P. 399, 16000 Alger (Algeria)], E-mail: mazrou_h@crna.dz

    2009-10-15

    The present work explores, through a comprehensive sensitivity study, a new methodology to find a suitable artificial neural network architecture which improves its performances capabilities in predicting two significant parameters in safety assessment i.e. the multiplication factor k{sub eff} and the fuel powers peaks P{sub max} of the benchmark 10 MW IAEA LEU core research reactor. The performances under consideration were the improvement of network predictions during the validation process and the speed up of computational time during the training phase. To reach this objective, we took benefit from Neural Network MATLAB Toolbox to carry out a widespread sensitivity study. Consequently, the speed up of several popular algorithms has been assessed during the training process. The comprehensive neural system was subsequently trained on different transfer functions, number of hidden neurons, levels of error and size of generalization corpus. Thus, using a personal computer with data created from preceding work, the final results obtained for the treated benchmark were improved in both network generalization phase and much more in computational time during the training process in comparison to the results obtained previously.

  15. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

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

  16. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  17. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  18. Kinetic parameters study based on burn-up for improving the performance of research reactor equilibrium core

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2014-01-01

    Full Text Available In this study kinetic parameters, effective delayed neutron fraction and prompt neutron generation time have been investigated at different burn-up stages for research reactor's equilibrium core utilizing low enriched uranium high density fuel (U3Si2-Al fuel with 4.8 g/cm3 of uranium. Results have been compared with reference operating core of Pakistan research Reactor-1. It was observed that by increasing fuel burn-up, effective delayed neutron fraction is decreased while prompt neutron generation time is increased. However, over all ratio beff/L is decreased with increasing burn-up. Prompt neutron generation time L in the understudy core is lower than reference operating core of reactor at all burn-up steps due to hard spectrum. It is observed that beff is larger in the understudy core than reference operating core of due to smaller size. Calculations were performed with the help of computer codes WIMSD/4 and CITATION.

  19. Power Control Method for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baang, Dane; Suh, Yongsuk; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Considering safety-oriented design concept and other control environment, we developed a simple controller that provides limiting function of power change- rate as well as fine tracking performance. The design result has been well-proven via simulation and actual application to a TRIGA-II type research reactor. The proposed controller is designed to track the PDM(Power Demand) from operator input as long as maintaining the power change rate lower than a certain value for stable reactor operation. A power control method for a TRIGA-II type research reactor has been designed, simulated, and applied to actual reactor. The control performance during commissioning test shows that the proposed controller provides fine control performance for various changes in reference values (PDM), even though there is large measurement noise from neutron detectors. The overshoot at low power level is acceptable in a sense of reactor operation.

  20. Quality control and performance evaluation of k{sub 0}-based neutron activation analysis at the Portuguese research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dung, H.M., E-mail: m.hodzung@itn.p [Instituto Technologico e Nuclear, Estrada Nacional 10, 2686-953 Sacavem (Portugal); Freitas, M.C. [Instituto Technologico e Nuclear, Estrada Nacional 10, 2686-953 Sacavem (Portugal); Blaauw, M. [Reactor Institute Delft, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands); Almeida, S.M.; Dionisio, I.; Canha, N.H. [Instituto Technologico e Nuclear, Estrada Nacional 10, 2686-953 Sacavem (Portugal)

    2010-10-11

    The quality control (QC) and performance evaluation for the k{sub 0}-based neutron activation analysis (k{sub 0}-NAA) at the Portuguese research reactor (RPI) has been developed with the intention of using the method to meet the demands of trace element analysis for the applications in environmental, epidemiological and nutritional studies amongst others. The QC and performance evaluation include the following aspects: (1) estimation of the overall/combined standard uncertainty from the primary uncertainty sources; (2) validation of the method using a synthetic multi-element standard (SMELS); and (3) analysis of the certified reference materials from the National Institute of Standards and Technology (USA): NIST-SRM-1633a and NIST-SRM-1648 and the reference material from the International Atomic Energy Agency: IAEA-RM-336, for the purpose of controlling the overall accuracy and precision of the analytical results. The obtained results revealed that the k{sub 0}-NAA method established at the RPI was fit for the purpose. The overall/combined standard uncertainty was estimated for elements of interest in the intended applications. The laboratory's analytical results as compared to the assigned values with the bias were less than 12% for most elements, except for a few elements which biased within 13-18%. The u-score values for most elements were less than |1.64|, except for Co, La and Ti within |1.64|-|1.96| and Sc, Cr, K and Sb within |1.96|-|2.58|. The NIST-1633a was also analyzed over 14 months for the purpose of evaluating the reproducibility of the method. The quality factors of k{sub 0}-NAA established at RPI were evaluated, proving that the method meets the requirements of trace element analysis, which is also considering the method's performance for which the k{sub 0}-NAA affords a specific, rapid and convenient capability for the intended applications.

  1. Strengthening IAEA Safeguards for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-01

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

  2. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  3. A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

    Science.gov (United States)

    Kamal, I.; Yunus, S. M.; Datta, T. K.; Zakaria, A. K. M.; Das, A. K.; Aktar, S.; Hossain, S.; Berliner, R.; Yelon, W. B.

    2016-07-01

    A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with a large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6mm thickness each. The monochromator design was optimized to provide maximum flux on 3mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30˚ (2θ) at each step and covers 120˚ in 4 steps. When the detector is placed at 1.6 m it subtends 20˚ at each step and covers 120˚ in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art technology

  4. Nuclear research reactors activities in INVAP

    Energy Technology Data Exchange (ETDEWEB)

    Ordonez, Juan Pablo [INVAP, Bariloche (Argentina)

    2013-07-01

    This presentation describes the different activities in the research reactor field that are being carried out by INVAP. INVAP is presently involved in the design of three new research reactors in three different countries. The RA-10 is a multipurpose reactor, in Argentina, planned as a replacement for the RA-3 reactor. INVAP was contracted by CNEA for carrying out the preliminary engineering for this reactor, and has recently been contracted by CNEA for the detailed engineering. CNEA groups are strongly involved in the design of this reactor. The RMB is a multipurpose reactor, planned by CNEN from Brazil. CNEN, through REDETEC, has contracted INVAP to carry out the preliminary engineering for this reactor. As the user requirements for RA-10 and RMB are very similar, an agreement was signed between Argentina and Brasil governments to cooperate in these two projects. The agreement included that both reactors would use the OPAL reactor in Australia, design and built by INVAP, as a reference reactor. INVAP has also designed the LPRR reactor for KACST in Saudi Arabia. The LPRR is a 30 kw reactor for educational purposes. KACST initially contracted INVAP for the engineering for this reactor and has recently signed the contract with INVAP for building the reactor. General details of these three reactors will be presented.

  5. Refurbishment of existing research reactors for BNCT

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F.E.; Gessaghi, V. [INVAP S.E., de Bariloche (Argentina)

    1997-12-01

    Some research reactors have been selected for the development of boron neutron capture therapy (BNCT) in the United States like the Massachusetts Institute of Technology research reactor, the University of Missouri research reactor 2 or the Brookhaven Medical Research Reactor, among others. These reactors have received excellent analyses and designs to accommodate extremely optimized beam shaping assemblies (BSAs) for the proper tuning of neutron spectra and absorption of undesired particles such as photons and fast neutrons. Due to the importance of BNCT in these facilities, the physicists and engineers have used many degrees of freedom for the optimization process.

  6. Fast Reactor Fuel Type and Reactor Safety Performance

    Energy Technology Data Exchange (ETDEWEB)

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and

  7. History and Actual State of Non-HEU Fission-Based Mo-99 Production with Low-Performance Research Reactors

    Directory of Open Access Journals (Sweden)

    S. Dittrich

    2013-01-01

    Full Text Available Fifty years ago, one of the worldwide first industrial production processes to produce fission-Mo-99 for medical use had been started at ZfK Rossendorf (now: HZDR, Germany. On the occasion of this anniversary, it is worth to mention that this original process (called LITEMOL now together with its target concept used at that time can still be applied. LITEMOL can be adapted very easily to various research reactors and applied at each site, which maybe still of interest for very small-scale producers. Besides this original process, two further and actually proven processes are suitable as well and recommended for small-scale LEU fission Mo-99 production also. They are known under the names KSA/KSS COMPACT and ROMOL LITE and will be described below.

  8. Improving the performance of the Egyptian second testing nuclear research reactor using interval type-2 fuzzy logic controller tuned by modified biogeography-based optimization

    Energy Technology Data Exchange (ETDEWEB)

    Sayed, M.M., E-mail: M.M.Sayed@ieee.org; Saad, M.S.; Emara, H.M.; Abou El-Zahab, E.E.

    2013-09-15

    Highlights: • A modified version of the BBO was proposed. • A novel method for interval type-2 FLC design tuned by MBBO was proposed. • The performance of the ETRR-2 was improved by using IT2FLC tuned by MBBO. -- Abstract: Power stabilization is a critical issue in nuclear reactors. The conventional proportional derivative (PD) controller is currently used in the Egyptian second testing research reactor (ETRR-2). In this paper, we propose a modified biogeography-based optimization (MBBO) algorithm to design the interval type-2 fuzzy logic controller (IT2FLC) to improve the performance of the Egyptian second testing research reactor (ETRR-2). Biogeography-based optimization (BBO) is a novel evolutionary algorithm that is based on the mathematical models of biogeography. Biogeography is the study of the geographical distribution of biological organisms. In the BBO model, problem solutions are represented as islands, and the sharing of features between solutions is represented as immigration and emigration between the islands. A modified version of the BBO is applied to design the IT2FLC to get the optimal parameters of the membership functions of the controller. We test the optimal IT2FLC obtained by modified biogeography-based optimization (MBBO) using the integral square error (ISE) and is compared with the currently used PD controller.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Usage of burnable poison on research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Villarino, Eduardo Anibal [INVAP S.E., San Carlos de Bariloche (Argentina)

    2002-07-01

    The fuel assemblies with burnable poison are widely used on power reactors, but there are not commonly used on research reactors. This paper shows a neutronic analysis of the advantages and disadvantages of the burnable poison usage on research reactors. This paper analyses both burnable poison design used on research reactors: Boron on the lateral wall and Cadmium wires. Both designs include a parametric study on the design parameters like the amount and geometry of the burnable poison. This paper presents the design flexibility using burnable poisons, it does not find an optimal or final design, which it will strongly depend on the core characteristics and fuel management strategy. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  13. Reactor containment research and development

    Energy Technology Data Exchange (ETDEWEB)

    Weil, N. A.

    1963-06-15

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

  14. Study on secondary shutdown systems in Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jalali, H.R.; Fadaei, A.H., E-mail: Fadaei_amir@aut.ac.ir; Gharib, M.

    2015-09-15

    Highlights: • A study was undertaken to summarize the techniques for secondary shutdown systems (SSS). • Neutronic calculation performed for proposed systems as SSS. • Dumping the heavy water stored in the reflector vessel is capable to shut down reactor. • Neutronic and transient calculation was done for validating the selected SSS. • All calculation shown that this system has advantages in safety and neutron economy. - Abstract: One important safety aspect of any research reactor is the ability to shut down the reactor. Usually, research reactors, currently in operation, have a single shutdown system based on the simultaneous insertion of the all control rods into the reactor core through gravity. Nevertheless, the International Atomic Energy Agency currently recommends use of two shutdown systems which are fully independent from each other to guarantee secure shutdown when one of them fails. This work presents an investigative study into secondary shutdown systems, which will be an important safety component in the research reactor and will provide another alternative way to shut down the reactor emergently. As part of this project, a study was undertaken to summarize the techniques that are currently used at world-wide research reactors for recognizing available techniques to consider in research reactors. Removal of the reflector, removal of the fuels, change in critical shape of reactor core and insertion of neutron absorber between the core and reflector are selected as possible techniques in mentioned function. In the next step, a comparison is performed for these methods from neutronic aspects. Then, chosen method is studied from the transient behavior point of view. Tehran research reactor which is a 5 MW open-pool reactor selected as a case study and all calculations are carried out for it. It has 5 control rods which serve the purpose of both reactivity control and shutdown of reactor under abnormal condition. Results indicated that heavy

  15. Gaseous fuel nuclear reactor research

    Science.gov (United States)

    Schwenk, F. C.; Thom, K.

    1975-01-01

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

  16. Dismantling decontamination of research reactor equipment

    Energy Technology Data Exchange (ETDEWEB)

    Voronik, N. I.; Davydov, Yu. P.; Shatilo, N. N. [Institute of Radioecological Problems Belarus Ac. Sci., Minsk-Sosny (Belarus)

    1999-07-01

    The purpose of the work was to check applicability of the existing and new compositions for decontamination and their adjustment to the specific conditions dealing with operation of the research reactor. (author)

  17. Supply of enriched uranium for research reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  18. Strengthening IAEA Safeguards for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-01

    During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half a dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors, for example, such characteristics include rapid on-line refueling, and a core design with room for such a large number of assemblies or targets that it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors, such as hot cells, where plutonium could be separated, could pose a safeguards challenge because, in some cases, they are not declared (because they are not located in the facility or because nuclear materials are not foreseen to be processed inside) and may not be accessible to inspectors in States without an Additional Protocol in force.

  19. In-reactor performance of pressure tubes in CANDU reactors

    Science.gov (United States)

    Rodgers, D. K.; Coleman, C. E.; Griffiths, M.; Bickel, G. A.; Theaker, J. R.; Muir, I.; Bahurmuz, A. A.; Lawrence, S. St.; Resta Levi, M.

    2008-12-01

    The pressure tubes in CANDU reactors have been operating for times up to about 25 years. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behaviour and discusses the factors controlling the behaviour of these components in currently operating CANDU reactors. The mechanical properties (such as ultimate tensile strength, UTS, and fracture toughness), and delayed-hydride-cracking properties (crack growth rate Vc, and threshold stress intensity factor, KIH) change with irradiation; the former reach a limiting value at a fluence of Pressure tubes exhibit elongation and diametral expansion. The deformation behaviour is a function of operating conditions and material properties that vary from tube-to-tube and as a function of axial location. Semi-empirical predictive models have been developed to describe the deformation response of average tubes as a function of operating conditions. For corrosion and, more importantly deuterium pickup, semi-empirical predictive models have also been developed to represent the behaviour of an average tube. The effect of material variability on corrosion behaviour is less well defined compared with other properties. Improvements in manufacturing have increased fracture resistance by minimising trace elements, especially H and Cl, and reduced variability by tightening controls on forming parameters, especially hot-working temperatures.

  20. Facility for a Low Power Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chalker, R. G.

    1949-09-14

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

  1. Performance Indicators of Operating Reactors

    Data.gov (United States)

    Nuclear Regulatory Commission — A list of Performance Indicators (PI) that are reported to the NRC by licensees at the end of each quarter in accordance with Inspection Manual Chapters (IMC) 0608,...

  2. Research on plasma core reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with 1-m-diam by 1-m-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-diam 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 UF/sub 6/ 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-cm/sup 3/ aluminum canister in the central region was fueled with UF/sub 6/ gas and fission density distributions determined. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

  3. Corrosion Minimization for Research Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Decommissioning of the Salaspils Research Reactor

    Directory of Open Access Journals (Sweden)

    Abramenkovs Andris

    2011-01-01

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

  5. Reactor pulse repeatability studies at the annular core research reactor

    Energy Technology Data Exchange (ETDEWEB)

    DePriest, K.R. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Trinh, T.Q. [Nuclear Facility Operations, Sandia National Laboratories, Mail Stop 0614, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Luker, S. M. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States)

    2011-07-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories is a water-moderated pool-type reactor designed for testing many types of objects in the pulse and steady-state mode of operations. Personnel at Sandia began working to improve the repeatability of pulse operations for experimenters in the facility. The ACRR has a unique UO{sub 2}-BeO fuel that makes the task of producing repeatable pulses difficult with the current operating procedure. The ACRR produces a significant quantity of photoneutrons through the {sup 9}Be({gamma}, n){sup 8}Be reaction in the fuel elements. The photoneutrons are the result of the gammas produced during fission and in fission product decay, so their production is very much dependent on the reactor power history and changes throughout the day/week of experiments in the facility. Because the photoneutrons interfere with the delayed-critical measurements required for accurate pulse reactivity prediction, a new operating procedure was created. The photoneutron effects at delayed critical are minimized when using the modified procedure. In addition, the pulse element removal time is standardized for all pulse operations with the modified procedure, and this produces less variation in reactivity removal times. (authors)

  6. Background Radiation Measurements at High Power Research Reactors

    CERN Document Server

    Ashenfelter, J; 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 $\\gamma$-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.

  7. Background radiation measurements at high power research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ashenfelter, J. [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Balantekin, B. [Department of Physics, University of Wisconsin, Madison, WI 53706 (United States); Baldenegro, C.X. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Band, H.R. [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Barclay, G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Bass, C.D. [Department of Chemistry and Physics, Le Moyne College, Syracuse, NY 13214 (United States); Berish, D. [Department of Physics, Temple University, Philadelphia, PA 19122 (United States); Bowden, N.S., E-mail: nbowden@llnl.gov [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Bryan, C.D. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Cherwinka, J.J. [Physical Sciences Laboratory, University of Wisconsin, Madison, WI 53706 (United States); Chu, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States); Classen, T. [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Davee, D. [Department of Physics, College of William and Mary, Williamsburg, VA 23187 (United States); Dean, D.; Deichert, G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Dolinski, M.J. [Department of Physics, Drexel University, Philadelphia, PA 19104 (United States); Dolph, J. [Physics Department, Brookhaven National Laboratory, Upton, NY 11973 (United States); Dwyer, D.A. [Physics Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Fan, S. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States); and others

    2016-01-11

    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.

  8. Model Based Cyber Security Analysis for Research Reactor Protection System

    Energy Technology Data Exchange (ETDEWEB)

    Sho, Jinsoo; Rahman, Khalil Ur; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Son, Hanseong [Joongbu Univ., Geumsan (Korea, Republic of)

    2013-07-01

    The study on the qualitative risk due to cyber-attacks into research reactors was performed using bayesian Network (BN). This was motivated to solve the issues of cyber security raised due to digitalization of instrumentation and control (I and C) system. As a demonstrative example, we chose the reactor protection system (RPS) of research reactors. Two scenarios of cyber-attacks on RPS were analyzed to develop mitigation measures against vulnerabilities. The one is the 'insertion of reactor trip' and the other is the 'scram halt'. The six mitigation measures are developed for five vulnerability for these scenarios by getting the risk information from BN.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. The rehabilitation/upgrading of Philippine Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-08-01

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

  13. Complete dismantling of the research reactor DIORIT

    Energy Technology Data Exchange (ETDEWEB)

    Beer, Hans-Frieder [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    2013-08-01

    The research reactor DIORIT at the Paul Scherrer Institute was a natural uranium reactor moderated by D{sub 2}O. It was put into operation in August 1960 and finally shut down in August 1977. The original dismantling plan, developed in 1980, comprised 3 phases and 13 steps. The dismantling started in 1982. It was interrupted for several times due to financial restrictions and during the last dismantling step due to the unexpected occurrence of asbestos. The dismantling could be successfully finished on September 11{sup th}, 2012. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    1987-11-01

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

  15. Neutron scattering at Australia's replacement research reactor

    Science.gov (United States)

    Robinson, R. A.; Kennedy, S. J.

    2002-01-01

    On August 25 1999, the Australian government gave final approval to build a research reactor to replace the existing HIFAR reactor at Lucas Heights. The replacement reactor, which will commence operation in 2005, will be multipurpose in function, with capabilities for both neutron-beam research and radioisotope production. Regarding beams, cold and thermal neutron sources are to be installed and the intent is to use supermirror guides, with coatings with critical angles up to 3 times that of natural Ni, to transport cold and thermal neutron beams into a large modern guide hall. The reactor and all the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP, SE and subcontractors in a turnkey contract. The goal is to have at least eight leading-edge neutron-beam instruments ready in 2005, and they will be developed by ANSTO and other contracted organisations, in consultation with the Australian user community and interested overseas parties. A review of the planned scientific capabilities, a description of the facility and a status report on the activities so far is given.

  16. The current status of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tri Wulan Tjiptono; Syarip

    1998-10-01

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

  17. Neutron flux optimization in irradiation channels at NUR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Meftah, B. [Division Reacteur, Centre de Recherche Nucleaire Draria (CRND), BP 43 Sebala DRARIA, Alger (Algeria)]. E-mail: b_meftah@yahoo.com; Zidi, T. [Division Reacteur, Centre de Recherche Nucleaire Draria (CRND), BP 43 Sebala DRARIA, Alger (Algeria); Bousbia-Salah, A. [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2 - 56126 Pisa (Italy)

    2006-09-15

    Optimization of neutron fluxes in experimental channels is of great concern in research reactor utilization. The general approach used at the NUR research reactor for neutron flux optimization in irradiation channels is presented. The approach is essentially based upon a judicious optimization of the core configuration combined with the improvement of reflector characteristics. The method allowed to increase the thermal neutron flux for radioisotope production purposes by more than 800%. Increases of up to 60% are also observed in levels of useful fluxes available for neutron diffraction experiments (small angle neutron scattering (SANS), neutron reflectometry, etc.). Such improvements in the neutronic characteristics of the NUR reactor opened new perspectives in terms of its utilization. More particularly, it is now possible to produce at industrial scales major radio-isotopes for medicine and industry and to perform, for the first time, material testing experiments. The cost of the irradiations in the optimized configuration is generally small when compared to those performed in the old configuration and an average reduction factor of about of 10 is expected in the case of production of Molybdenum-99 (isotope required for the manufacturing of Technetium-99 medical kits). In addition to these important results, safety analysis studies showed that the more symmetrical nature of the core geometry leads to a more adequately balanced reactivity control system and contributes quite efficiently to the operational safety of the NUR reactor. Results of comparisons between calculations and measurements for a series of parameters of importance in reactor operation and safety showed good agreement.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-30

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

  19. 钠冷快堆超导钠泵建模与性能研究%Modeling and Performance Research on Sodium Cooled Fast Reactor Superconductive Sodium Pump

    Institute of Scientific and Technical Information of China (English)

    杨志达; 赵佳宁; 韩伟实

    2014-01-01

    为克服钠冷快堆机械式钠泵机械磨损、噪声大、泄漏以及普通电磁泵流量小、扬程小等问题,提出了大流量鞍型磁体的电磁泵作为驱动钠循环的主泵设计方案。对其结构进行了研究并建立了相应的数学模型,利用M atlab编写程序进行了不同电流、磁感应强度和温度条件下泵的扬程、流量和效率的性能研究。结果表明,扬程随通道宽度的增加、电流的减小、磁感应强度的减小而减小,效率随电流的减小、磁感应强度的减小而减小,温度高于400℃时由于接触电阻的降低可使效率提高,鞍型超导钠泵的流量可由电流、磁感应强度控制,但较为实用的是电流控制。该研究可为具体设计提供依据。%In order to overcome the shortcoming of sodium cooled fast reactor mechanical sodium pump including mechanical wear ,large noise ,sodium leakage and the shortcom-ing of electromagnetic pump including small flow rate and small pump head ,the large flow rate sodium pump design based on saddle superconductive magnet was put forward . The structure of sodium pump was researched and the mathematic model was estab-lished .The performance including pump head ,flow rate and efficiency were researched at different currents ,magnetic flux densities and temperatures using the computer code programmed with Matlab .The result shows that the pump head will reduce with the channel width increasing ,current reducing and magnetic flux density reducing ,and the efficiency will reduce with the current reducing and magnetic flux density reducing .If the temperature overtops 400 ℃ ,the efficiency will be high because of contact resistance disappearing .T he flow rate of the saddle type superconductive sodium pump can be controlled by current and magnetic flux density ,and the current is more suitable .This research can provide foundation for concreting design .

  20. Detailed Burnup Calculations for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Leszczynski, F. [Centro Atomico Bariloche (CNEA), 8400 S. C. de Bariloche (Argentina)

    2011-07-01

    A general method (RRMCQ) has been developed by introducing a microscopic burn up scheme which uses the Monte Carlo calculated spatial power distribution of a research reactor core and a depletion code for burn up calculations, as a basis for solving nuclide material balance equations for each spatial region in which the system is divided. Continuous energy dependent cross-section libraries and full 3D geometry of the system is input for the calculations. The resulting predictions for the system at successive burn up time steps are thus based on a calculation route where both geometry and cross-sections are accurately represented, without geometry simplifications and with continuous energy data. The main advantage of this method over the classical deterministic methods currently used is that RRMCQ System is a direct 3D method without the limitations and errors introduced on the homogenization of geometry and condensation of energy of deterministic methods. The Monte Carlo and burn up codes adopted until now are the widely used MCNP5 and ORIGEN2 codes, but other codes can be used also. For using this method, there is a need of a well-known set of nuclear data for isotopes involved in burn up chains, including burnable poisons, fission products and actinides. For fixing the data to be included on this set, a study of the present status of nuclear data is performed, as part of the development of RRMCQ method. This study begins with a review of the available cross-section data of isotopes involved in burn up chains for research nuclear reactors. The main data needs for burn up calculations are neutron cross-sections, decay constants, branching ratios, fission energy and yields. The present work includes results of selected experimental benchmarks and conclusions about the sensitivity of different sets of cross-section data for burn up calculations, using some of the main available evaluated nuclear data files. Basically, the RRMCQ detailed burn up method includes four

  1. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  2. A novel concept for CRIEC-driven subcritical research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nieto, M.; Miley, G.H. [Illinois Univ., Fusion Studies Lab., Dept. of Nuclear, Plasma, and Radiological Engineering, Urbana, IL (United States)

    2001-07-01

    A novel scheme is proposed to drive a low-power subcritical fuel assembly by means of a long Cylindrical Radially-convergent Inertial Electrostatic Confinement (CRIEC) used as a neutron source. The concept is inherently safe in the sense that the fuel assembly remains subcritical at all times. Previous work has been done for the possible implementation of CRIEC as a subcritical assembly driver for power reactors. However, it has been found that the present technology and stage of development of IEC-based neutron sources can not meet the neutron flux requirements to drive a system as big as a power reactor. Nevertheless, smaller systems, such as research and training reactors, could be successfully driven with levels of neutron flux that seem more reasonable to be achieved in the near future by IEC devices. The need for custom-made expensive nuclear fission fuel, as in the case of the TRIGA reactors, is eliminated, and the CRIEC presents substantial advantages with respect to the accelerator-driven subcritical reactors in terms of simplicity and cost. In the present paper, a conceptual design for a research/training CRIEC-driven subcritical assembly is presented, emphasizing the description, principle of operation and performance of the CRIEC neutron source, highlighting its advantages and discussing some key issues that require study for the implementation of this concept. (author)

  3. Safe operation and maintenance of research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Munsorn, S. [Reactor Operation Division, Office of Atomic Energy for Peace, Chatuchak, Bangkok (Thailand)

    1999-10-01

    The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U{sub 3}O{sub 8}- Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

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

    CERN Document Server

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

    2016-01-01

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

  5. The AFR. An approved network of research reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  6. Convective cooling in a pool-type research reactor

    Science.gov (United States)

    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.

  7. Convective cooling in a pool-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sipaun, Susan, E-mail: susan@nm.gov.my [Malaysian Nuclear Agency, Industrial Technology Division, Blok 29T, Bangi 43200, Selangor (Malaysia); Usman, Shoaib, E-mail: usmans@mst.edu [Missouri University of Science and Technology, Nuclear Engineering, 222 Fulton Hall 301 W.14th St., Rolla 64509 MO (United States)

    2016-01-22

    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 (U{sub 3}Si{sub 2}Al) 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{sup −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{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

  8. Remediation of Site of Decommissioning Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Danilovich, A.S.; Ivanov, O.P.; Lemus, A.V.; Pavlenko, V.I.; Potapov, V.N.; Semenov, S.G.; Shisha, A.D.; Chesnokov, A.V. [National Research Center ' Kurchatov Institute' , 123182, Moscow (Russian Federation)

    2014-07-01

    In the world the most widespread method of soil decontamination consists of removing the contaminated upper layer and sending it for long-term controlled storage. However, implementation of this soil cleanup method for remediation of large contaminated areas would involve high material and financial expenditures, because it produces large amounts of radioactive waste demanding removal to special storage sites. Contaminated soil extraction and cleanup performed right on the spot of remediation activities represents a more advanced and economically acceptable method. Radiological separation of the radioactive soil allows reducing of amount of radwaste. Studies performed during the liquidation of the Chernobyl accident consequences revealed that a considerable fraction of radioactivity is accumulated in minute soil grains. So, the separation of contaminated soil by size fractions makes it possible to extract and concentrate the major share of radioactivity in the fine fraction. Based on these researches water gravity separation technology was proposed by Bochvar Institute. The method extracts the fine fraction from contaminated soil. Studies carried out by Bochvar Institute experts showed that, together with the fine fraction (amounting to 10-20% of the initial soil), this technology can remove up to 85-90% of contaminating radionuclides. The resulting 'dirty' soil fraction could be packaged into containers and removed as radwaste, and decontaminated fractions returned back to their extraction site. Use of radiological and water gravity separations consequently increases the productivity of decontamination facility. Efficiency of this technology applied for contaminated soil cleanup was confirmed in the course of remediation of the contaminated territories near decommissioning research reactor in the Kurchatov Institute. For soil cleaning purposes, a special facility implementing the technology of water gravity separation and radiometric monitoring of soil

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  10. Research reactor de-fueling and fuel shipment

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Design and performance of subgrade biogeochemical reactors.

    Science.gov (United States)

    Gamlin, Jeff; Downey, Doug; Shearer, Brad; Favara, Paul

    2017-02-18

    Subgrade biogeochemical reactors (SBGRs), also commonly referred to as in situ bioreactors, are a unique technology for treatment of contaminant source areas and groundwater plume hot spots. SBGRs have most commonly been configured for enhanced reductive dechlorination (ERD) applications for chlorinated solvent treatment. However, they have also been designed for other contaminant classes using alternative treatment media. The SBGR technology typically consists of removal of contaminated soil via excavation or large-diameter augers, and backfill of the soil void with gravel and treatment amendments tailored to the target contaminant(s). In most cases SBGRs include installation of infiltration piping and a low-flow pumping system (typically solar-powered) to recirculate contaminated groundwater through the SBGR for treatment. SBGRs have been constructed in multiple configurations, including designs capable of meeting limited access restrictions at heavily industrialized sites, and at sites with restrictions on surface disturbance due to sensitive species or habitat issues. Typical performance results for ERD applications include 85 to 90 percent total molar reduction of chlorinated volatile organic compounds (CVOCs) near the SBGR and rapid clean-up of adjacent dissolved contaminant source areas. Based on a review of the literature and CH2M's field-scale results from over a dozen SBGRs with a least one year of performance data, important site-specific design considerations include: 1) hydraulic residence time should be long enough for sufficient treatment but not too long to create depressed pH and stagnant conditions (e.g., typically between 10 and 60 days), 2) reactor material should balance appropriate organic mulch as optimal bacterial growth media along with other organic additives that provide bioavailable organic carbon, 3) a variety of native bacteria are important to the treatment process, and 4) biologically mediated generation of iron sulfides along with

  12. Monte Carlo modelling of TRIGA research reactor

    Science.gov (United States)

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

    2010-10-01

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

  13. A study on the establishment of component/equipment performance criteria considering Heavy Water Reactor characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Keun Sun; Kwon, Young Chul; Lee, Min Kyu; Lee, Yun Soo [Sunmoon Univ., Asan (Korea, Republic of); Chang, Seong Hoong; Ryo, Chang Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Kim, Soong Pyung; Hwnag, Jung Rye; Chung, Chul Kee [Chosun Univ., Gwangju (Korea, Republic of)

    2002-03-15

    Foreign and domestic technology trends, regulatory requirements, design and researches for heavy water reactors are analyzed. Safety design guides of Canada industry and regulatory documents and consultative documents of Canada regulatory agency are reviewed. Applicability of MOST guidance 16 Revision 'guidance for technical criteria of nuclear reactor facility' is reviewed. Specific performance criteria are established for components and facilities for heavy water reactor.

  14. Architecture dependent availability analysis of RPS for Research Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Rahman, Khalilur; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Son, Hanseong [Joongbu Univ., Geumsan (Korea, Republic of); Kim, Youngki; Park, Jaekwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The research reactors are categorized into two broad categories, Low power research reactors and medium to high power research reactors. According to IAEA TECDOC-1234, Research reactors with 0.250- 2.0 MW power rating or 2.5-10 Χ 10{sup 11} η/cm{sup 2}. s flux are termed low power reactor whereas research reactors ranging from 2-10 MW power rating or 0.1-10 Χ 10{sup 13} η/cm{sup 2}. s are considered as Medium to High power research reactors. Some other standards (IAEA NP-T-5.1) define multipurpose research reactor ranging from power few hundred KW to 10 MW as low power research reactor. The aim of this research, in this article, was to identify a configuration of architecture which gives highest availability with maintaining low cost of manufacturing. In this regard, two configurations of a single channel of RPS are formulated in the current article and their fault trees were developed using AIMS PSA software to get the unavailability. This is a starting point of attempt towards the standardization of I and C architecture for low and medium power research reactors.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-15

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

  16. High Performance Photocatalytic Oxidation Reactor System Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Pioneer Astronautics proposes a technology program for the development of an innovative photocatalytic oxidation reactor for the removal and mineralization of...

  17. Overcoming the effects of stress on reactor operator performance

    Energy Technology Data Exchange (ETDEWEB)

    He Xuhong; Wei Li; Zhao Bingquan [Tsinghua Univ., Nuclear Power Plant Simulation Training Center, Beijing (China)

    2003-03-01

    Reactor operators may be exposed to significant levels of stress during plant emergencies and their performance may be affected by the stress. This paper first identified the potential sources of stress in the nuclear power plant, then discussed the ways in which stress is likely to affect the reactor operators, and finally identified several training approaches for reducing or eliminating stress effects. The challenges for effective stress reducing training may seem daunting, yet the challenges are real and must be addressed. This paper reviewed researches in training design, knowledge and skill acquisition, and training transfer point to a number of strategies that can be used to address these challenges and lead to more effective training and development. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-10

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

  19. The current status of nuclear research reactor in Thailand

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  20. Meteodiffusive Characterization of Algiers' Nuclear Research Reactor

    Directory of Open Access Journals (Sweden)

    Mourad Messaci

    2007-01-01

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

  1. Design characteristics and requirements of irradiation holes for research reactor experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, B. C.; Chae, H. T.; Lee, C. S.; Seo, C. G

    2003-07-01

    In order to be helpful for the design of a new research reactor with high performance, are summarized the applications of research reactors in various fields and the design characteristics of experimental facility such as vertical irradiation holes and beam tubes. Basic requirements of such experimental facilities are also described. Research reactor has been widely utilized in various fields such as industry, engineering, medicine, life science, environment etc., and now the application fields are gradually being expanded together with the development of technology. Looking into the research reactors which are recently constructed or in plan, it seems that to develop a multi-purpose research reactor with intensive neutron beam research capability has become tendency. In the layout of the experimental facilities, the number and configuration of irradiation and beam holes should be optimized to meet required test conditions such as neutron flux at the early design stage. But, basically high neutron flux is required to perform experiments efficiently. In this aspect, neutron flux is regarded as one of important parameters to judge the degree of research reactor performance. One of main information for a new research reactor design is utilization demands and requirements of experimental holes. So basic requirements which should be considered in a new research reactor design were summarized from the survey of experimental facilities characteristics of various research reactors with around 20 MW thermal power and the experiences of HANARO utilization. Also is suggested an example of the requirements of experimental holes such as size, number and neutron flux, which are thought as minimum, in a new research reactor for exporting to developing countries such as Vietnam.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  3. Oxidation performance of graphite material in reactors

    Institute of Scientific and Technical Information of China (English)

    Xiaowei LUO; Xinli YU; Suyuan YU

    2008-01-01

    Graphite is used as a structural material and moderator for high temperature gas-cooled reactors (HTGR). When a reactor is in operation, graphite oxida-tion influences the safety and operation of the reactor because of the impurities in the coolant and/or the acci-dent conditions, such as water ingress and air ingress. In this paper, the graphite oxidation process is introduced, factors influencing graphite oxidation are analyzed and discussed, and some new directions for further study are pointed out.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-02-17

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

  5. Research Reactor Design for Export to Myanmar

    Energy Technology Data Exchange (ETDEWEB)

    Win Naing, Lay Lay Myint and Myung-Hyun Kim [Kyunghee Univ. Yongin (Korea, Republic of)

    2006-07-01

    Myanmar is striving to acquire the innovative technology in all field areas including maritime, aerospace and nuclear engineering. There is a high intention to construct a new research reactor for peaceful purposes. The Ministry of Science and Technology (MOST) and Ministry of Education (MOE) are the important government organizations for Myanmar's education and they control most of institutes, universities and colleges. The Department of Atomic Energy (DAE), one of the departments under MOST, leads research projects such as for radiation protection as well as radiation application and coordinates government departments and institutions regarding nuclear energy and its applications. Myanmar's Scientific and Technological Research Department (MSTRD) under MOST guides researches in metallurgy, polymer, pharmacy and biotechnology and so on, and acts as an official body for Myanmar industrial standard. The Department of Higher Education (DHE) under MOE controls art and science universities and colleges including research centers such as Asia Research Center (ARC), Universities Research Center (URC), Microbiology Research Center and so on and does to expand research areas and to utilize advanced technology in science. The wide use of radiation and radioisotopes is developed in Myanmar especially for the field areas such as Medical Science and Agricultural Science. Co{sup 60}, I{sup 131} and Tc{sup 99} are the major use of radioisotopes in diagnosis and therapy. In Agricultural Science, H{sup 3}, C{sup 14}, C{sup 60} etc are used to provide biological effects of radiations on plants, radio-isotopic study of soil physics and tracer studies.

  6. Performance of pressure tubes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-20

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Feasibility of Thermoelectric Waste Heat Recovery from Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  11. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  12. Technical Research for Dedicated Isotope Production Reactor of South Africa

    Institute of Scientific and Technical Information of China (English)

    ZOU; Yao; LIU; Xing-min; CHEN; Hui-qiang; SUN; Zhen; WU; Yuan-yuan

    2012-01-01

    <正>Research reactor plays an important part in nuclear science and technology, application and power development. Currently, many countries in Middle East and Africa are ready to develop their own nuclear industry. South Africa sent its User Requirements Specification (URS) for a dedicated isotope production reactor to several institutes or companies, among of which Department of Reactor Engineering Research and Design (DRERD) in China Institute of Atomic Energy (CIAE) is a competitive candidate.

  13. Initial decommissioning planning for the Budapest research reactor

    OpenAIRE

    Toth Gabor

    2011-01-01

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

  14. Initial decommissioning planning for the Budapest research reactor

    Directory of Open Access Journals (Sweden)

    Toth Gabor

    2011-01-01

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

  15. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Baldev Raj

    2009-06-01

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

  16. Decommissioning Project for the Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, U. S.; Park, J. H.; Paik, S. T. (and others)

    2009-02-15

    In 2008, tried to complete the whole decommissioning project of KRR-1 and KRR-2 and preparing work for memorial museum of KRR-1 reactor. Now the project is delayed for 3 months because of finding unexpected soil contamination around facility and treatment of. To do final residual radioactivity assessment applied by MARSSIM procedure. Accumulated decommissioning experiences and technologies will be very usefully to do decommissioning other nuclear related facility. At the decommissioning site of the uranium conversion plant, the decontamination of the dismantled carbon steel waste are being performed and the lagoon 1 sludge waste is being treated this year. The technologies and experiences obtained from the UCP dismantling works are expected to apply to other fuel cycle facilities decommissioning. The lagoon sludge treatment technology is the first applied technology in the actual field and it is expected that this technology could be applied to other country.

  17. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  18. Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Worrall, Andrew [ORNL; Todosow, Michael [Brookhaven National Laboratory (BNL)

    2016-01-01

    Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include: increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle performance

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-01

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

  20. Status of reactor shielding research in the United States

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-15

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

  2. Development of a decommissioning strategy for the MR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bylkin, Boris; Gorlinsky, Yury; Kolyadin, Vyacheslav; Pavlenko, Vitaly [RRC Kurchatov Institute, Moscow (Russian Federation); Craig, David; Fecitt, Lorna [NUKEM Limited, Dounreay (United Kingdom); Harman, Neil; Jackson, Roger [Serco Technical and Assurance Services, Warrington (United Kingdom); Lobach, Yury [Inst. for Nuclear Research of NASU, Kiev (Ukraine)

    2010-03-15

    A description of the selected decommissioning strategy for the research reactor MR at the site of the Kurchatov Institute in Moscow is presented. The MR reactor hall is planned to be used as a temporary fuel store for the other research reactors on the site. On the basis of the site-specific conditions and over-all decommissioning goals, it was identified that phased immediate decommissioning is the preferable option. The current status of the reactor, expected final conditions and the sequence of decommissioning works are shown. (orig.)

  3. Proceedings of the sixth Asian symposium on research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-08-01

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

  4. Reactor performances and microbial communities of biogas reactors: effects of inoculum sources.

    Science.gov (United States)

    Han, Sheng; Liu, Yafeng; Zhang, Shicheng; Luo, Gang

    2016-01-01

    Anaerobic digestion is a very complex process that is mediated by various microorganisms, and the understanding of the microbial community assembly and its corresponding function is critical in order to better control the anaerobic process. The present study investigated the effect of different inocula on the microbial community assembly in biogas reactors treating cellulose with various inocula, and three parallel biogas reactors with the same inoculum were also operated in order to reveal the reproducibility of both microbial communities and functions of the biogas reactors. The results showed that the biogas production, volatile fatty acid (VFA) concentrations, and pH were different for the biogas reactors with different inocula, and different steady-state microbial community patterns were also obtained in different biogas reactors as reflected by Bray-Curtis similarity matrices and taxonomic classification. It indicated that inoculum played an important role in shaping the microbial communities of biogas reactor in the present study, and the microbial community assembly in biogas reactor did not follow the niche-based ecology theory. Furthermore, it was found that the microbial communities and reactor performances of parallel biogas reactors with the same inoculum were different, which could be explained by the neutral-based ecology theory and stochastic factors should played important roles in the microbial community assembly in the biogas reactors. The Bray-Curtis similarity matrices analysis suggested that inoculum affected more on the microbial community assembly compared to stochastic factors, since the samples with different inocula had lower similarity (10-20 %) compared to the samples from the parallel biogas reactors (30 %).

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

    Science.gov (United States)

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

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  7. 10 MW research reactor simulation using fuel plate type

    Energy Technology Data Exchange (ETDEWEB)

    Mustafa, M. El Sayed, E-mail: memmm67@yahoo.com [Reactors Department, Nuclear Researches Center, Inshas (Egypt); Shaat, M. [Reactors Department, Nuclear Researches Center, Inshas (Egypt); Kady, M. El [Mechanical Power Engineering Department, Faculty of Engineering, Al Azhar University, Cairo (Egypt)

    2016-04-15

    A computer code was established named ET-RR-1-10 to investigate the thermal hydraulic behavior of the ETRR1 (first Egyptian research reactor) research reactor when its power upgraded to 10 MW using the new fuel plate elements type. The work done include both normal and flow reduction conditions. The code modeled primary loop, secondary lop, and reactor kinetics. All code models used finite difference technique. The code results were tested against the available corresponding experimental data taken from a similar research reactor MITR (Massachusetts Institute of Technology research reactor) for the sake of code validation. The results showed good agreement, and the code can be used for thermal hydraulic calculations.

  8. Licensing of the Australian replacement research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Garea, Veronica B. [INVAP S.E., San Carlos de Bariloche (Argentina)

    2002-07-01

    The Australian Nuclear Science and Technology Organisation (ANSTO)'s Replacement Research Reactor has been submitted to a comprehensive licensing process of which peer review has been a fundamental part. Following Australian Regulation, an application for a site licence was the first step supported by an Environmental Impact Statement approved by The Minister for the Environment and Heritage, and a Reference Accident Analysis. After the site licence had been granted and the contract awarded to the Designer and Constructor, INVAP S.E:, a 2500 page Preliminary Safety Analysis Report was submitted by ANSTO to the Australian Radiation Protection and Nuclear Safety Agency (ARPANSA), which conducted its review. ARPANSA requested that the PSAR be also reviewed by an experts mission from the International Atomic Energy Agency. The PSAR was also reviewed by the Argentine Regulatory Body, it was submitted to public examination in Australia and it was reviewed by international experts hired as consultants by several Australian organisations. A public forum was also held in Sydney. The Regulator, the applicant and the Designer-Constructor maintained constant interaction during the whole process, so that questions, comments and observations that arose from the review of the PSAR were fed back to the designers. This process allowed for a robust, safe design enriched by the results of the safety analysis and review process. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  10. Progress with OPAL, the new Australian research reactor

    Indian Academy of Sciences (India)

    R A Robinson

    2008-11-01

    Australian science is entering a new `golden age', with the start-up of bright new neutron and photon sources in Sydney and Melbourne, in 2006 and 2007 respectively. The OPAL reactor and the Australian Synchrotron can be considered as the greatest single investment in scientific infrastructure in Australia's history. They will essentially be `sister' facilities, with a common open user ethos, and a vision to play a major role in international science. Fuel was loaded into the reactor in August 2006, and full power was (20 MW) achieved in November 2006. The first call for proposals was made in 2007, and commissioning experiments have taken place well before then. The first three instruments in operation are high-resolution powder diffractometer (for materials discovery), high-intensity powder diffractometer (for kinetics experiments and small samples) and a strain scanner (for mechanical engineering and industrial applications). These are closely followed by four more instruments with broad application in nanoscience, condensed matter physics and other scientific disciplines. Instrument performance will be competitive with the best research-reactor facilities anywhere. To date there is committed funding for nine instruments, with a capacity to install a total of ∼ 18 beamlines. An update will be given on the status of OPAL, its thermal and cold neutron sources, its instruments and the first results.

  11. Reactor pressure vessel structural integrity research

    Energy Technology Data Exchange (ETDEWEB)

    Pennell, W.E.; Corwin, W.R. [Oak Ridge National Lab., TN (United States)

    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. Interim irradiated fuel storage facility for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lolich, Jose [INVAP SE, Bariloche (Argentina)

    2002-07-01

    In most research reactors irradiated fuel discharged from the reactor is initially stored underwater inside the reactor building for along period of time. This allows for heat dissipation and fission product decay. In most cases this initial storage is done in a irradiated fuel storage facility pool located closed to the reactor core. After a certain cooling time, the fuel discharged should be relocated for long-term interim storage in a Irradiated Fuel Storage (IFS) Facility. IFS facilities are required for the safe storage of irradiated nuclear fuel before it is reprocessed or conditioned for disposal as radioactive waste. The IFS Facility described in this report is not an integral part of an operating nuclear reactor. This facility many be either co-located with nuclear facilities (such as a nuclear reactor or reprocessing plant) or sited independently of other nuclear facilities. (author)

  13. Transient two-phase performance of LOFT reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Chen, T.H.; Modro, S.M.

    1983-01-01

    Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed.

  14. Modeling of operating history of the research nuclear reactor

    Science.gov (United States)

    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.

  15. Need for space-time analyses of research reactor transients

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F.E. [INVAP S.E., de Bariloche (Argentina)

    1997-12-01

    The success of the point-kinetics approximation to represent the time behavior of research reactors relies on the fact that research reactor cores are small enough to be neutronically tightly coupled; the core is small when measured in diffusion lengths. This fact implies that a certain change in a part of the core is immediately observed by the whole system. The propagation of changes is so fast that the core exhibits a shape function that is practically unchanged during the transient; the amplitude function, the only unknown of the problem, represents the full knowledge of the core response. One is immediately warned to look for the truth of this assumption. How small should a research reactor core be to be sure that point kinetics is a valid assumption? This question is becoming increasingly important because the tendency is to increase the size of research reactor cores to make them capable of various simultaneous uses (multipurpose characteristics), with powers in the range of tens of thermal megawatts. One of the lines of investigation at the Department of Reactor Physics is related to scenarios of Materials Test Reactor (MTR)-type research reactor transients for which space-time kinetics would bring a more profound insight than point kinetics.

  16. Role of research reactors for nuclear power program in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Soentono, S.; Arbie, B. [National Atomic Energy Agency, Batan (Indonesia)

    1994-12-31

    The main objectives of nuclear development program in Indonesia are to master nuclear science and technology, as well as to utilise peaceful uses of nuclear know-how, aiming at stepwisely socioeconomic development. A Triga Mark II, previously of 250 kW, reactor in Bandung has been in operation since 1965 and its design power has been increased to 1000 kW in 1972. Using core grid of the Triga 250 kW, BATAN designed and constructed the Kartini Reactor in Yogyakarta which started its operation in 1979. Both of these Triga reactors have served a wide spectrum of utilisation, such as training of manpower in nuclear engineering as well as radiochemistry, isotope production and beam research activities in solid state physics. In order to support the nuclear power development program in general and to suffice the reactor experiments further, simultaneously meeting the ever increasing demand for radioisotope, the third reactor, a multipurpose reactor of 30 MW called GA. Siwabessy (RSG-GAS) has been in operation since 1987 at Serpong near Jakarta. Each of these reactors has strong cooperation with Universities, namely the Bandung Institute of Technology at Bandung, the Gadjah Mada University at Yogyakarta, and the Indonesia University at Jakarta and has facilitated the man power development required. The role of these reactors, especially the multipurpose GA. Siwabessy reactor, as essential tools in nuclear power program are described including the experience gained during preproject, construction and commissioning, as well as through their operation, maintenance and utilisation.

  17. China Advanced Research Reactor Project Progress in 2011

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    2011, China Advanced Research Reactor (CARR) Project finished the B stage commissioning and resolved the relative technical problems. Meanwhile, the acceptance items and the cold neutron source were carrying out.

  18. Conceptual Study for development of a low power research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, C.; Kim, H. S.; Park, J. H.; Chae, H. T.; Lee, B. C. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    Even though the nuclear society is again facing with difficult situations after Fukusima accident, some countries still continues to consider nuclear power as one option of national energy sources and to introduce nuclear energy. As a research reactor has been regarded as a step-stone to establish infrastructures for the nuclear power development program, some countries that have plan to introduce the nuclear power energy are considering to construct a research reactor. Particularly, a low power research reactor whose main purpose is basic researches on the nuclear technology and education/training would be of interest to developing countries when taking the economy and level of science and technology into consideration. And many low power research reactors at operation are obsolescent and their numbers are decreasing. Hence, some concepts on a low power research reactor are being studied for the future needs. This paper presents the conceptual study on the basic requirements and the preliminary design features of a low power research reactor.

  19. PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, L.; Diamond, D.; Xu, J.; Carew, J.; Rorer, D.

    2004-03-31

    Detailed reactor physics and 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 analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron and photon transport calculations were performed with the MCNP code to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim safety arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim safety arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the

  20. Renewing Liquid Fueled Molten Salt Reactor Research and Development

    Science.gov (United States)

    Towell, Rusty; NEXT Lab Team

    2016-09-01

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

  1. Development of an educational nuclear research reactor simulator

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

    This paper introduces the development of a research reactor educational simulator based on LabVIEW that allows the training of operators and studying different accident scenarios and the effects of operational parameters on the reactor behavior. Using this simulator, the trainee can test the interaction between the input parameters and the reactor activities. The LabVIEW acts as an engine implements the reactor mathematical models. In addition, it is used as a tool for implementing the animated graphical user interface. This simulator provides the training requirements for both of the reactor staff and the nuclear engineering students. Therefore, it uses dynamic animation to enhance learning and interest for a trainee on real system problems and provides better visual effects, improved communications, and higher interest levels. The benefits of conducting such projects are to develop the expertise in this field and save costs of both operators training and simulation courses.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  3. Research and development on next generation reactor (phase I)

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-10-01

    The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system reactor according to design margin decrease. (Author).

  4. Oak Ridge National Laboratory Research Reactor Experimenters' Guide

    Energy Technology Data Exchange (ETDEWEB)

    Cagle, C.D. (comp.)

    1982-04-01

    The Oak Ridge National Laboratory (ORNL) operates six research reactors dedicated to research and development work as well as radioisotope production. These reactors are used by ORNL and qualified non-ORNL research and development groups. The purpose of this report is to provide information to research personnel concerning the facilities and the ORNL research and services groups available to assist in the design, fabrication, operation, and post-operation examination of irradiation assemblies. Safety and operability reviews and quality assurance requirements are also described.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  6. Activities for extending the lifetime of MINT research reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  7. Neutron beam facilities at the Australian Replacement Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kennedy, Shane; Robinson, Robert; Hunter, Brett [Physics Division, ANSTO, Lucas Heights (Australia)

    2001-03-01

    Australia is building a research reactor to replace the HIFAR reactor at Lucas Heights by the end of 2005. Like HIFAR, the Replacement Research Reactor will be multipurpose with capabilities for both neutron beam research and radioisotope production. It will be a pool-type reactor with thermal neutron flux (unperturbed) of 4 x 10{sup 14} n/cm{sup 2}/sec and a liquid D{sub 2} cold neutron source. Cold and thermal neutron beams for neutron beam research will be provided at the reactor face and in a large neutron guide hall. Supermirror neutron guides will transport cold and thermal neutrons to the guide hall. The reactor and the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP S.E. under contract. The neutron beam instruments will be developed by ANSTO, in consultation with the Australian user community. This status report includes a review the planned scientific capabilities, a description of the facility and a summary of progress to date. (author)

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

    Science.gov (United States)

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

    2017-02-01

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

  9. Neutron scattering at Australia's replacement research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kennedy, S.J.; Robinson, R.A.; Hunter, B.A. [Physics Division, ANSTO, Lucas Heights, NSW (Australia)

    2001-03-01

    On August 25{sup th} 1999, the Australian government gave final approval to build a research reactor to replace the existing HIFAR reactor at Lucas Heights. The replacement reactor, which will commence operation in 2005, will be multipurpose with capabilities for both neutron beam research and radioisotope production. Cold, and thermal neutron sources are to be installed and supermirror guides will transport cold and thermal neutron beams into a large modern guide hall. The reactor and associated infrastructure is to be built by INVAP, SE and subcontractors under contract. The neutron beam instruments will be developed by ANSTO in consultation with the Australian user community and interested overseas parties. We review the planned scientific capabilities, give a description of the facility and a status report on the activities so far. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  11. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  12. Status and some safety philosophies of the China advanced research reactor CARR

    Energy Technology Data Exchange (ETDEWEB)

    Luzheng Yuan [China Inst. of Atomic Energy, Beijing, BJ (China). Reactor Engineering Research and Design Dept.

    2001-07-01

    The existing two research reactors, HWRR (heavy water research reactor) and SPR (swimming pool reactor), have been operated by China Institute of Atomic Energy (CIAE) since, respectively, 1958 and 1964, and are both in extending service and facing the aging problem. It is expected that they will be out of service successively in the beginning decade of the 21{sup st} century. A new, high performance and multipurpose research reactor called China advanced research reactor (CARR) will replace these two reactors. This new reactor adopts the concept of inverse neutron trap compact core structure with light water as coolant and heavy water as the outer reflector. Its design goal is as follows: under the nuclear power of 60MW, the maximum unperturbed thermal neutron flux in peripheral D{sub 2}O reflector not less than 8 x 10{sup 14} n/cm{sup 2}. s while in central experimental channel, if the central cell to be replaced by an experimental channel, the corresponding value not less than 1 x 10{sup 15} n/cm{sup 2}. s. The main applications for this research reactor will cover RI production, neutron scattering experiments, NAA and its applications, neutron photography, NTD for monocrystaline silicon and applications on reactor engineering technology. By the end of 1999, the preliminary design of CARR was completed, then the draft of preliminary safety analysis report (PSAR) was submitted to the relevant authority at the end of 2000 for being reviewed. Now, the CARR project has entered the detail design phase and safety reviewing procedure for obtaining the construction permit from the relevant licensing authority. This paper will only briefly introduce some aspects of safety philosophy of CARR design and PSAR. (orig.)

  13. Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)

    2013-07-01

    A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

  14. Design and Construction of Operation Bridge for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Kwangsub; Choi, Jinbok; Lee, Jongmin; Oh, Jinho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The operation bridge contains a lower working deck mounted on a saddle that travels on rails. Upright members are mounted on the saddle to support the upper structure and two hoist monorails. The saddle contains an anti-derail system that is composed of seismic lugs and guide rollers. The operation bridge travels along the rails to transport the fuel assembly, irradiated object, and reactor components in the pools by using tools. Hoists are installed at the top girder. The hoist is suspended from the monorail by means of a motor driven trolley that runs along the monorail. Movements of hoist and trolley are controlled by using the control pendant switch. Processes of design and construction of the operation bridge for the research reactor are introduced. The operation bridge is designed under consideration of functions of handling equipment in the pool and operational limits for safety. Structural analysis is carried out to evaluate the structural integrity in the seismic events. Tests and inspections are also performed during fabrication and installation to confirm the function and safety of the operation bridge.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-01

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

  16. Industrial Research Performance Management

    CERN Document Server

    Samsonowa, Tatjana

    2012-01-01

    Managers are increasingly concerned with the typical methods available for organizational performance measurement and control. Research into performance measurement, within the field of innovation management, has been variously approached through frameworks for performance measurement in general (for example, the Balanced Scorecard by Norton and Kaplan), R&D performance management, and surveys on in-use Key Performance Indicators (KPIs). It is striking, however, that almost no research has focused explicitly on the performance measurement of research activities, or indeed tried to develop a sy

  17. The present situations and perspectives on utilization of research reactors in Thailand

    Science.gov (United States)

    Chongkum, Somporn

    2002-01-01

    The Thai Research Reactor 1/Modification 1, a TRIGA Mark III reactor, went critical on November 7, 1977. It has been playing a central role in the development of both Office of Atomic Energy for Peace (OAEP) and nuclear application in Thailand. It has a maximum power of 2 MW (thermal) at steady state and a pulsing capacity of 2000 MW. The highest thermal neutron flux at a central thimber is 1×10 13 n/cm 2/s, which is extensively utilized for radioisotope production, neutron activation analysis and neutron beam experiments, i.e. neutron scattering, prompt gamma analysis and neutron radiography. Following the nuclear technological development, the OAEP is in the process of establishing the Ongkharak Nuclear Research Center (ONRC). The center is being built in Nakhon Nayok province, 60 km northeast of Bangkok. The centerpiece of the ONRC is a multipurpose 10 MW TRIGA research reactor. Facilities are included for the production of radioisotopes for medicine, industry and agriculture, neutron transmutation doping of silicon, and neutron capture therapy. The neutron beam facilities will also be utilized for applied research and technology development as well as training in reactor operations, performance of experiments and reactor physics. This paper describes a recent program of utilization as well as a new research reactor for enlarging the perspectives of its utilization in the future.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  19. Operating manual for the Health Physics Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1985-11-01

    This manual is intended to serve as a guide in the operation and maintenance of the Health Physics Researh Reactor (HPRR) of the Health Physics Dosimetry Applications Research (DOSAR) Facility. It includes descriptions of the HPRR and of associated equipment such as the reactor positioning devises and the derrick. Procedures for routine operation of the HPRR are given in detail, and checklists for the various steps are provided where applicable. Emergency procedures are similarly covered, and maintenance schedules are outlined. Also, a bibliography of references giving more detailed information on the DOSAR Facility is included. Changes to this manual will be approved by at least two of the following senior staff members: (1) the Operations Division Director, (2) the Reactor Operations Department Head, (3) the Supervisor of Reactor Operations TSF-HPRR Areas. The master copy and the copy of the manual issued to the HPRR Operations Supervisor will always reflect the latest revision. 22 figs.

  20. A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Jacques V Hugo; David I Gertman; Jeffrey C Joe

    2014-08-01

    This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

  1. Neoliberalism, Performativity and Research

    Science.gov (United States)

    Roberts, Peter

    2007-07-01

    This paper provides a critical analysis of New Zealand's Performance Based Research Fund (PBRF). The first section sketches the development and implementation of the PBRF. The second section evaluates the scheme, concentrating on three themes: the relationship between privatization, competition and research performance; the standardization of research; and motivations for research. The paper acknowledges the thorough work completed by the Tertiary Education Advisory Commission and other policy groups in laying the foundation for the adoption of performance-based research funding in New Zealand. It is argued, however, that when viewed in its larger context, the PBRF constitutes a continuation of neoliberal trends already well established in New Zealand's tertiary education system.

  2. The performance of structured packings in trickle-bed reactors.

    NARCIS (Netherlands)

    Frank, M.J.W.; Frank, M.J.W.; Kuipers, J.A.M.; Versteeg, Geert; van Swaaij, Willibrordus Petrus Maria

    1999-01-01

    An experimental study was carried out to investigate whether the use of structured packings might improve the mass transfer characteristics and the catalyst effectiveness of a trickle-bed reactor. Therefore, the performances of a structured packing, consisting of KATAPAK elements, and a dumped

  3. The performance of structured packings in trickle-bed reactors.

    NARCIS (Netherlands)

    Frank, M.J.W.; Kuipers, J.A.M.; Versteeg, G.F.; Swaaij, van W.P.M.

    1999-01-01

    An experimental study was carried out to investigate whether the use of structured packings might improve the mass transfer characteristics and the catalyst effectiveness of a trickle-bed reactor. Therefore, the performances of a structured packing, consisting of KATAPAK elements, and a dumped packi

  4. The Performance of Structured Packings in Trickle-Bed Reactors

    NARCIS (Netherlands)

    Frank, M.J.W.; Kuipers, J.A.M.; Versteeg, G.F.; Swaaij, W.P.M. van

    1999-01-01

    An experimental study was carried out to investigate whether the use of structured packings might improve the mass transfer characteristics and the catalyst effectiveness of a trickle-bed reactor. Therefore, the performances of a structured packing, consisting of KATAPAK elements, and a dumped packi

  5. The Jules Horowitz Reactor - A new High Performance European Material Testing Reactor open to International Users Present Status and Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Iracane, Daniel; Bignan, Gilles [CEA Atomic Energy Commission Saclay Batiment 121- 91191 Gif Sur Yvette (France); Lindbaeck, Jan-Erik; Blomgren, Jan [VATTENFALL AB Nuclear Power Jaemtlandsgatan 99 SE-16287 Stockholm (Sweden)

    2010-07-01

    The development of sustainable nuclear energy requires R and D on fuel and material behaviour under irradiation with a high level of performance in order to meet the needs and challenges for the benefit of industry, research and public bodies. These stakes require a sustainable and secured access to an up-to-date high performance Material Testing Reactor. Following a broad survey within the European Research Area, the international community agreed that the need for Material Test Reactors in support of nuclear power plant safety and operation will continue in the context of sustainable nuclear energy. The Jules Horowitz Reactor project (JHR) copes with this context. JHR is designed as a user facility addressing the needs of the international community. This means: - flexibility with irradiation loops able to reproduce a large variation in operation conditions of different power reactor technologies, - high flux capacity to address Generations II, III, and IV needs. JHR is designed, built and operated as an international user facility because: - Given the maturity and globalization of the industry, domestic tools have no more the required level of economic and technical efficiency. Meanwhile, countries with nuclear energy need an access to high performance irradiation experimental capabilities to support technical skill and guarantee the competitiveness and safety of nuclear energy. - Many research items related to safety or public policy (waste management, etc.) require international cooperation to share costs and benefits of resulting consensus. JHR design is optimised for offering high performance material and fuel irradiation capability for the coming decades. This project is driven and funded by an international consortium gathering vendors, utilities and public stakeholders. This consortium has been set up in March 2007 when the construction began. The construction is in progress and the start of operation is scheduled for 2014. The JHR is a research

  6. EFFECT OF DYE CONCENTRATION ON SEQUENCING BATCH REACTOR PERFORMANCE

    Directory of Open Access Journals (Sweden)

    A. A. Vaigan ، M. R. Alavi Moghaddam ، H. Hashemi

    2009-01-01

    Full Text Available Reactive dyes have been identified as problematic compounds in textile industries wastewater as they are water soluble and cannot be easily removed by conventional aerobic biological treatment systems. The treatability of a reactive dye (Brill Blue KN-R by sequencing batch reactor and the influence of the dye concentration on system performance were investigated in this study. Brill Blue KN-R is one of the main dyes that are used in textile industries in Iran. Four cylindrical Plexiglas reactors were run for 36 days (5 days for acclimatization of sludge and 31 days for normal operation at different initial dye concentrations. The dye concentrations were adjusted to be 20, 25, 30 and 40 mg/L in the reactors R1, R2, R3 and R4, respectively. In all reactors, effective volume, influent wastewater flowrate and sludge retention time were 5.5 L, 3.0 L/d and 10 d, respectively. According to the obtained data, average dye removal efficiencies of R1, R2, R3 and R4 were 57% ± 2, 50.18% ± 3, 44.97% ± 3 and 30.98% ± 3, respectively. The average COD removal efficiencies of all reactors were 97% ± 1, 97.12% ± 1, 96.93% ± 1 and 97.22% ± 1, respectively. The dye removal efficiency was decreased by increasing the dye concentration with the correlation coefficient of 0.997.

  7. Analysis on Electromagnetic Characteristics of Research Reactor Control Rod Drive Mechanism for Thrust Force Improvement

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hyung; Choi, Myoung Hwan; Yu, Je Yong; Cho, Yeong Garp; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The control rod drive mechanism (CRDM) is the part of reactor regulating system (RRS), which is located in the reactor pool top or the room below the reactor pool. The function of the CRDM is to insert, withdraw or maintain neutron absorbing material (control rod) at any required position within the reactor core, in order to the reactivity of the core. There are so many kinds of CRDM, such as magnetic-jack type, hydraulic type, rack and pinion type, chain type and linear or rotary step motor and so on. As a part of a new project, we are investigating the movable coil electromagnetic drive mechanism (MCEDM) which is new scheme for the reactor control rod adopted by China Advanced Research Reactor (CARR). To have a better knowledge of the electromagnetic and magnetic characteristics, numerical models of MCEDM are proposed. Especially in order to achieve improved thrust force, numerical magnetic field calculations for various kinds of magnetic and electromagnetic configuration have been performed. As a result, we present the improved design of MCEDM for research reactor

  8. BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Williamson, Richard L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Novascone, Stephen Rhead [Idaho National Lab. (INL), Idaho Falls, ID (United States); Medvedev, Pavel G. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on the formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.

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

    Science.gov (United States)

    Radulescu, Laura; Pavelescu, Margarit

    2010-01-01

    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.

  10. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1988-05-01

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

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

    Science.gov (United States)

    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.

  12. Prevention of ageing of research reactors by design

    Energy Technology Data Exchange (ETDEWEB)

    Boado, J. [INVAP, Bariloche (Argentina); Lolich, J. [INVAP, Bariloche (Argentina)

    1995-12-31

    However, it is our experience as designers and builders of research reactors, that the most important cause of ageing of any experimental installation, is the loss of motivation of the personnel involved in the operation and maintenance, when the objectives for the utilisation of the facility change or research programs are abandoned for whatever reason. We therefore have endeavoured to design and construct research reactors with several engineering features and with an untraditional approach to the training of the future operator of the facility. During all phases of the design and construction of the reactor, we develop in the future operator of the facility the capacity, not only to operate it properly, but to innovate and to adapt the installation to the daily operating problems due to new requirements and options that might not have been foreseen when the facility was ordered. The versatility of the operator is thus a further guarantee against ageing by abandonment. (orig./HP)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

  17. Hafnium as a neutron absorber in research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F.E.; Villarino, E.A. [INVAP S.E., S.C. de Bariloche (Argentina)

    1995-12-31

    Research reactors usually require the periodic replacement of control rods because of the burnup of the neutron absorber material, even if devices have been mainly withdrawn. The purpose of this paper is the study of the reactivity worth of hafnium as a neutron absorber in the multi-purpose-type research reactor (MPTRR). The MPTRR design develops 20 MW, and it is configured by five rows of six low-enriched uranium (200% enrichment) materials test reactor fuels inside a Zircaloy chimney. The standard control plates are given by two rows of three plates of silver-indium-cadmium alloy (85% silver, 10% indium, and 5% cadmium, in weight percents; 0.26 cm of maximum absorber width), separating the central three rows of fuel elements from the peripheral first and fifth rows. The burnup of control plates decreases the reactivity worth below the licensing criteria in {approximately}6 months.

  18. Status and future of the WWR-M research reactor in Kiev

    Energy Technology Data Exchange (ETDEWEB)

    Bazavov, D.A.; Gavrilyuk, V.I.; Kirischuk, V.I.; Kochetkov, V.V.; Lysenko, M.V.; Makarovskiy, V.N.; Scherbachenko, A.M.; Shevel, V.N.; Slisenko, V.I. [Institute for Nuclear Research, Kiev (Ukraine)

    2001-07-01

    Kiev WWR-M Research Reactor, operated at maximum power of 10 MW, was put into operation in 1960 and during its 40-years history has been used to perform numerous studies in different areas of science and technology. Due to a number of technical problems the Research Reactor, the only one in Ukraine, was shut down in 1993 and then put into operation in 1999 again. Now there is an intention to reconstruct Kiev Research Reactor. The upgraded Research Reactor would allow solving such problems as the safe operation of Ukrainian NPPs, radioisotope production and, naturally, fundamental and applied research. The main problem for the successful operation of Kiev Research Reactor is the management and storage of spent fuel at the site, since after core unloading the spent fuel storage appears to be practically completed. So it is absolutely necessary to ship the most part of the spent fuel for reprocessing and as soon as possible. Besides, there is a need to build up the new spent fuel storage, because the tank of available storage requires careful inspection for corrosion. (author)

  19. Progress in the decommissioning planning for the Kiev’s research reactor WWR-M

    Directory of Open Access Journals (Sweden)

    Lobach Yuri N.

    2010-01-01

    Full Text Available The Kiev’s research reactor WWR-M has been in operation for more than 50 years and its further operation is planned for no less than 8-10 years. The acting nuclear legislation of Ukraine demands from the operator to perform the decommissioning planning during the reactor operation stage as early as possible. Recently, the Decommissioning Program has been approved by the regulatory body. The Program is based on the plans for the further use of the reactor site and foresees the strategy of immediate dismantling. The Program covers the whole de- commissioning process and represents the main guiding document during the whole decommissioning period, which determines and substantiates the principal technical and organizational activities on the preparation and implementation of the reactor decommissioning, the consequence of the decommissioning stages, the sequence of planned works and measures as well as the necessary conditions and infrastructure for the provision and safe implementation.

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

    CERN Document Server

    Michiels, Ianthe

    2016-01-01

    In the past decades, various nuclear reactor neutrino experiments have measured a deficit in the flux of antineutrinos coming from the reactor at short reactor-detector distances, when compared to theoretical calculations. One of the experiments designed to investigate this reactor antineutrino anomaly is the SoLid experiment. It uses the compact BR2 research reactor from the SCK-CEN in Mol, Belgium, to perform reactor antineutrino flux measurements at very short baseline. These proceedings discuss the general detection concepts of the SoLid experiment and its novel detector technology. The performance of the SoLid design is demonstrated with some results of the analysis of the data gathered with the experiment's first large scale test module, SM1.

  1. China Advanced Research Reactor Project Progress in 2012

    Institute of Scientific and Technical Information of China (English)

    ZHAO; Tie-jun

    2012-01-01

    <正>In 2012, all the commissioning for the China Advanced Research Reactor (CARR) had been finished and the diffraction pattern had been successfully obtained on the neutron scattering spectrometer. Meanwhile, the cold neutron source project and the acceptance items of CARR project had been carrying out.

  2. Kartini Research Reactor prospective studies for neutron scattering application

    Energy Technology Data Exchange (ETDEWEB)

    Widarto [Yogyakarta Nuclear Research Center, BATAN (Indonesia)

    1999-10-01

    The Kartini Research Reactor (KRR) is located in Yogyakarta Nuclear Research Center, Yogyakarta - Indonesia. The reactor is operated for 100 kW thermal power used for research, experiments and training of nuclear technology. There are 4 beam ports and 1 column thermal are available at the reactor. Those beam ports have thermal neutron flux around 10{sup 7} n/cm{sup 2}s each other and used for sub critical assembly, neutron radiography studies and Neutron Activation Analysis (NAA). Design of neutron collimator has been done for piercing radial beam port and the calculation result of collimated neutron flux is around 10{sup 9} n/cm{sup 2}s. This paper describes experiment facilities and parameters of the Kartini research reactor, and further more the prospective studies for neutron scattering application. The purpose of this paper is to optimize in utilization of the beam ports facilities and enhance the manpower specialty. The special characteristic of the beam ports and preliminary studies, pre activities regarding with neutron scattering studies for KKR is presented. (author)

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

    CERN Document Server

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

    2002-01-01

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

  4. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

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

  5. Human reliability analysis of the Tehran research reactor using the SPAR-H method

    Directory of Open Access Journals (Sweden)

    Barati Ramin

    2012-01-01

    Full Text Available The purpose of this paper is to cover human reliability analysis of the Tehran research reactor using an appropriate method for the representation of human failure probabilities. In the present work, the technique for human error rate prediction and standardized plant analysis risk-human reliability methods have been utilized to quantify different categories of human errors, applied extensively to nuclear power plants. Human reliability analysis is, indeed, an integral and significant part of probabilistic safety analysis studies, without it probabilistic safety analysis would not be a systematic and complete representation of actual plant risks. In addition, possible human errors in research reactors constitute a significant part of the associated risk of such installations and including them in a probabilistic safety analysis for such facilities is a complicated issue. Standardized plant analysis risk-human can be used to address these concerns; it is a well-documented and systematic human reliability analysis system with tables for human performance choices prepared in consultation with experts in the domain. In this method, performance shaping factors are selected via tables, human action dependencies are accounted for, and the method is well designed for the intended use. In this study, in consultations with reactor operators, human errors are identified and adequate performance shaping factors are assigned to produce proper human failure probabilities. Our importance analysis has revealed that human action contained in the possibility of an external object falling on the reactor core are the most significant human errors concerning the Tehran research reactor to be considered in reactor emergency operating procedures and operator training programs aimed at improving reactor safety.

  6. Fuel shuffling optimization for the Delft research reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Use of research and test reactors for SPD development and calibration

    Energy Technology Data Exchange (ETDEWEB)

    LaFontaine, M.W.R. [Physics Solutions Inc., Kitchener, ON (Canada)

    2011-07-01

    Prior to using a research or test reactor for performance studies or calibration of self powered detectors, it is first necessary to fully characterize the reactor environment in the region to be utilized. This presentation details Characterization Experiments performed to quantify research/test reactor core/site parameters as they would apply for use with SPD applications. Methods will be described to: Determine the Westcott parameter, r (T {sub n}/T {sub o}) , for the region of interest; Characterize the neutron energy spectrum in terms of the cadmium absorption cut-off, i.e., consider neutrons of energy 5kT < 0.13 eV to be thermal neutrons, and neutrons of energy 5kT > 0.13 eV to be epithermal neutrons; Determine T {sub n}, the effective neutron temperature, in the region of interest; Determine the gamma flux in the region of interest; and, Establish SPD calibration standard detectors.

  9. Electronic imaging system for neutron radiography at a low power research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, F.J.O., E-mail: fferreira@ien.gov.b [Instituto de Engenharia Nuclear, Comissao Nacional de Energia Nuclear, Caixa Postal 68550, CEP 21945-970, Rio de Janeiro (Brazil); Silva, A.X.; Crispim, V.R. [PEN/COPPE-DNC/POLI CT, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970 Rio de Janeiro (Brazil)

    2010-08-15

    This paper describes an electronic imaging system for producing real time neutron radiography from a low power research reactor, which will allow inspections of samples with high efficiency, in terms of measuring time and result analysis. This system has been implanted because of its potential use in various scientific and industrial areas where neutron radiography with photographic film could not be applied. This real time system is installed in neutron radiography facility of Argonauta nuclear research reactor, at the Instituto de Engenharia Nuclear of the Comissao Nacional de Energia Nuclear, in Brazil. It is adequate to perform real time neutron radiography of static and dynamic events of samples.

  10. Operational limitations of light water reactors relating to fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, H S

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed.

  11. Operational limitations of light water reactors relating to fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, H S

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed.

  12. Design of a New Research Reactor: Preliminary Conceptual Design (3rd Year)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, B. C.; Chae, H. T. and others

    2006-01-15

    A research reactor design is a kind of integral engineering project and a process to obtain a concrete shape through several years of concept development, conceptual design, basic design and detail design. So it requires close cooperation in various areas as well as lots of manpower and cost. The overall process at each stage may be said to be similar except for some stage-specific works. In 2005 as last year of a concept development stage, investigations on the various concepts of the fuel, reactor structure and systems which can meet the requirements established. The requirements for the process systems and I and C systems have also been embodied. The major tasks planned at the early of 2005 have been performed for each area of reactor design as follows: Establishment of the fuel and reactor core concept, and the core analysis, Preliminary thermal-hydraulic and safety analyses for the conceptual cores, Establishment and improvement of analysis system, Concept developments of the reactor structures and major systems, Test and test plan to verify the developed concepts, International cooperation to establish the foundations for exporting a research reactor.

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

    Science.gov (United States)

    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.

  14. Experience of IEA-R1 research reactor spent fuel transportation back to United States

    Energy Technology Data Exchange (ETDEWEB)

    Frajndlich, Roberto [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div. de Operacao do Reator IEAR-R1m]. E-mail: frajndli@net.ipen.br; Perrotta, Jose A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div.de Engenharia do Nucleo]. E-mail: perrotta@net.ipen.br; Maiorino, Jose Rubens [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Diretoria de Reatores]. E-mail: maiorino@net.ipen.br; Soares, Adalberto Jose [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Dept. de Reatores]. E-mail: ajsoares@net.ipen.br

    1998-07-01

    IPEN/CNEN-SP is sending the IEA-R1 Research Reactor spent fuels from USA origin back to this country. This paper describes the experience in organizing the negotiations, documents and activities to perform the transport. Subjects as cask licensing, transport licensing and fuel failure criteria for transportation are presented. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-13

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

  16. Current status of neutron activation analysis in HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Yong Sam; Moon, Jong Hwa; Sohn, Jae Min [Korea Atomic Energy Research Institute, Daejeon (Korea)

    2003-03-01

    The facilities for neutron activation analysis in the HANARO (Hi-flux Advanced Neutron Application Research Reactor) are described and the main applications of NAA (Neutron Activation Analysis) are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system were installed at three irradiation holes of HANARO at the end of 1995. The performance of the NAA facility was examined to identify the characteristics of the tube transfer system, irradiation sites and custom-made polyethylene irradiation capsule. The available thermal neutron fluxes at irradiation sites are in the range of 3 x 10{sup 13} - 1 x 10{sup 14} n/cm{sup 2}{center_dot}s and cadmium ratios are in 15 - 250. For an automatic sample changer for gamma-ray counting, a domestic product was designed and manufactured. An integrated computer program (Labview) to analyse the content was developed. In 2001, PGNAA (Prompt Gamma Neutron Activation Analysis) facility has been installed using a diffracted neutron beam of ST1. NAA has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials, and various polymers for research and development. The improvement of analytical procedures and establishment of an analytical quality control and assurance system were studied. Applied research and development for the environment, industry and human health by NAA and its standardization were carried out. For the application of the KOLAS (Korea Laboratory Accreditation Scheme), evaluation of measurement uncertainty and proficiency testing of reference materials were performed. Also to verify the reliability and to validate analytical results, intercomparison studies between laboratories were carried out. (author)

  17. Research on Precaution and Detection Technology for Flow Blockage of Plate-type Fuel Element in Research Reactors

    Institute of Scientific and Technical Information of China (English)

    DING; Li; QIAO; Ya-xin; ZHANG; Nian-peng; LUO; Bei-bei; HUA; Xiao; JIA; Shu-jie; YAN; Hui-yang

    2013-01-01

    The main aim of this study is to offer the technical support for safety operation and management of research reactors using plate-type fuel assemblies in China,which is performed from analysis of precaution measures for flow blockage and detection methods of accidents.Study shows that most accidents were induced by in-core foreign objects and the swelling of fuel

  18. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  19. Impact of reactor water chemistry on cladding performance

    Energy Technology Data Exchange (ETDEWEB)

    Cox, B. [University of Toronto, Centre for Nuclear Engineering, Toronto, Ontario (Canada)

    1997-07-01

    Water chemistry may have a major impact on fuel cladding performance in PWRs. If the saturation temperature on the surface of fuel cladding is exceeded, either because of the thermal hydraulics of the system, or because of crud deposition, then LiOH concentration can occur within thick porous oxide films on the cladding. This can degrade the protective film and accelerate the corrosion rate of the cladding. If sufficient boric acid is also present in the coolant then these effects may be mitigated. This is normally the case through most of any reactor fuel cycle. Extensive surface boiling may disrupt this equilibrium because of the volatility of boric acid in steam. Under such conditions severe cladding corrosion can ensue. The potential for such effects on high burnup cladding in CANDU reactors, where bone acid is not present in the primary coolant, is discussed. (author)

  20. Computational Neutronics Methods and Transmutation Performance Analyses for Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    R. Ferrer; M. Asgari; S. Bays; B. Forget

    2007-03-01

    The once-through fuel cycle strategy in the United States for the past six decades has resulted in an accumulation of Light Water Reactor (LWR) Spent Nuclear Fuel (SNF). This SNF contains considerable amounts of transuranic (TRU) elements that limit the volumetric capacity of the current planned repository strategy. A possible way of maximizing the volumetric utilization of the repository is to separate the TRU from the LWR SNF through a process such as UREX+1a, and convert it into fuel for a fast-spectrum Advanced Burner Reactor (ABR). The key advantage in this scenario is the assumption that recycling of TRU in the ABR (through pyroprocessing or some other approach), along with a low capture-to-fission probability in the fast reactor’s high-energy neutron spectrum, can effectively decrease the decay heat and toxicity of the waste being sent to the repository. The decay heat and toxicity reduction can thus minimize the need for multiple repositories. This report summarizes the work performed by the fuel cycle analysis group at the Idaho National Laboratory (INL) to establish the specific technical capability for performing fast reactor fuel cycle analysis and its application to a high-priority ABR concept. The high-priority ABR conceptual design selected is a metallic-fueled, 1000 MWth SuperPRISM (S-PRISM)-based ABR with a conversion ratio of 0.5. Results from the analysis showed excellent agreement with reference values. The independent model was subsequently used to study the effects of excluding curium from the transuranic (TRU) external feed coming from the LWR SNF and recycling the curium produced by the fast reactor itself through pyroprocessing. Current studies to be published this year focus on analyzing the effects of different separation strategies as well as heterogeneous TRU target systems.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-04-01

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

  2. Characterization of Novel Calorimeters in the Annular Core Research Reactor *

    Directory of Open Access Journals (Sweden)

    Hehr Brian D.

    2016-01-01

    Full Text Available 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

    planned to create fully equipped complex facility possible to perform various experiments on the intensive neutron beam. Epithermal neutron beam enables development across the full spectrum of materials research for example shielding concrete tests or electronic devices construction improvement. Due to recent reports on the construction of the accelerator for the Boron Neutron Capture Therapy (BNCT) it has the opportunity to become useful and successful method in the fight against brain and other types of cancers not treated with well known medical methods. In Europe there is no such epithermal neutron source which could be used throughout the year for training and research for scientist working on BNCT what makes the stand unique in Europe. Also our research group which specializes in mixed radiation dosimetry around nuclear and medical facilities would be able to carry out research on new detectors and methods of measurements for radiological protection and in-beam (therapeutic) dosimetry. Another group of scientists from National Centre for Nuclear Research, where MARIA research reactor is located, is involved in research of gamma detector systems. There is an idea to develop Prompt-gamma Single Photon Emission Computed Tomography (Pg- SPECT). This method could be used as imaging system for compounds emitting gamma rays after nuclear reaction with thermal neutrons e.g. for boron concentration in BNCT. Inside the room, where H2 channel is located, there is another horizontal channel - H1 which is also unused. Simultaneously with the construction of the H2 stand it will be possible to create special pneumatic horizontal mail inside the H1 channel for irradiation material samples in the vicinity of the core i.e. in the distal part of the H1 channel. It might expand the scope of research at the planned neutron station. Secondly it is planned to equip both stands with moveable positioning system, video system and facilities to perform animal experiments (anaesthesia, vital

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-10-01

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

  5. 75 FR 27368 - Aerotest Operations, Inc., Aerotest Radiography and Research Reactor; Notice of Consideration of...

    Science.gov (United States)

    2010-05-14

    ... COMMISSION Aerotest Operations, Inc., Aerotest Radiography and Research Reactor; Notice of Consideration of... INFORMATION CONTACT: Cindy Montgomery, Project Manager, Research and Test Reactors Licensing Branch, Division... Operating License No. R-98 for the Aerotest Radiography and Research Reactor (ARRR), currently held by...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  7. Safety culture and quality management of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

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

  9. [Effects of the size of magnetic particles of immobilized enzyme reactors on the digestion performance].

    Science.gov (United States)

    Zhang, Jiao; Zhou, Lianqi; Tian, Fang; Zhang, Yangjun; Qian, Xiaohong

    2013-02-01

    We applied immobilized enzyme reactors prepared with different sizes of magnetic particles into protein and proteome digestion. In addition, the influences of different sizes of the magnetic particles were studied on the reunion, enzyme efficiency and leakage sites. The experimental results showed that in comparison with the submicron magnetic particles, the amount of trypsin immobilized on the magnetic nanoparticles was 3. 5 times more than that of the submicron magnetic particles. However, the enzymatic efficiency was at the same level when the same amount of trypsin was used, and the reunion phenomenon was obviously improved when the size of the magnetic nanoparticles increased. Taking the immobilized enzyme reactor of 20 nm magnetic nanoparticles as an example, the digestion performance was further examined. The experimental results showed that rapid digestion could be achieved within 1 mm when the mass ratio of the trypsin and bovine serum albumin was 1:1. The peptide number of 0 missed cleavage site and the sequence coverage changed little after the protein was digested for 10 mm. It was concluded that the digestion efficiency of the immobilized enzyme reactor was much better than that of the in-solution digestion. When the immobilized enzyme reactors and the free trypsin were used for digestion, little differences of the leakage sites were found. Therefore, the immobilized enzyme reactors prepared with different sizes of magnetic particles can be applied in proteomic research for quick and efficient digestion.

  10. On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

    Energy Technology Data Exchange (ETDEWEB)

    Chauvin, J.P. [CEA,DEN, DER, SPEX, Experimental Physics Service, Cadarache, F-13108 St-Paul-Lez-Durance (France); Blaise, P. [CEA, DEN, DER, SPEX Experimental Programs Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France)

    2015-07-01

    The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physics calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)

  11. Preparation Before Signature of Upgrade of Algeria Heavy Water Research Reactor Contract

    Institute of Scientific and Technical Information of China (English)

    LI; Song; ZAN; Huai-qi; XU; Qi-guo; JIA; Yu-wen

    2012-01-01

    <正>Algeria heavy water research reactor (Birine) is a multiple-purpose research reactor, which was constructed with the help of China more than 20 years ago. By request of Algeria, China will upgrade the research reactor; so as to improve the status of current reactor such as equipment ageing, shortage of spare parts, several systems do not meet requirements of current standards and criteria etc.

  12. Titer-plate formatted continuous flow thermal reactors: Design and performance of a nanoliter reactor.

    Science.gov (United States)

    Chen, Pin-Chuan; Park, Daniel S; You, Byoung-Hee; Kim, Namwon; Park, Taehyun; Soper, Steven A; Nikitopoulos, Dimitris E; Murphy, Michael C

    2010-08-06

    Arrays of continuous flow thermal reactors were designed, configured, and fabricated in a 96-device (12 × 8) titer-plate format with overall dimensions of 120 mm × 96 mm, with each reactor confined to a 8 mm × 8 mm footprint. To demonstrate the potential, individual 20-cycle (740 nL) and 25-cycle (990 nL) reactors were used to perform the continuous flow polymerase chain reaction (CFPCR) for amplification of DNA fragments of different lengths. Since thermal isolation of the required temperature zones was essential for optimal biochemical reactions, three finite element models, executed with ANSYS (v. 11.0, Canonsburg, PA), were used to characterize the thermal performance and guide system design: (1) a single device to determine the dimensions of the thermal management structures; (2) a single CFPCR device within an 8 mm × 8 mm area to evaluate the integrity of the thermostatic zones; and (3) a single, straight microchannel representing a single loop of the spiral CFPCR device, accounting for all of the heat transfer modes, to determine whether the PCR cocktail was exposed to the proper temperature cycling. In prior work on larger footprint devices, simple grooves between temperature zones provided sufficient thermal resistance between zones. For the small footprint reactor array, 0.4 mm wide and 1.2 mm high fins were necessary within the groove to cool the PCR cocktail efficiently, with a temperature gradient of 15.8°C/mm, as it flowed from the denaturation zone to the renaturation zone. With temperature tolerance bands of ±2°C defined about the nominal temperatures, more than 72.5% of the microchannel length was located within the desired temperature bands. The residence time of the PCR cocktail in each temperature zone decreased and the transition times between zones increased at higher PCR cocktail flow velocities, leading to less time for the amplification reactions. Experiments demonstrated the performance of the CFPCR devices as a function of flow

  13. Characterization of radioactive aerosols in Tehran research reactor containment

    Directory of Open Access Journals (Sweden)

    Moradi Gholamreza

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

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

  15. Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Foyto, L [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Kutikkad, K [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; McKibben, J C [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Peters, N. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Stevens, J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2012-09-01

    This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.

  16. Developing strategic plans for effective utilization of research reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

  17. Design requirement for electrical system of an advanced research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, H. K.; Kim, Y. K.; Wu, J. S.; Ryu, J. S

    2004-12-01

    An advanced research reactor is being designed since 2002 and the conceptual design has been completed this year for the several types of core. Also the fuel was designed for the potential cores. But the process system, the I and C system, and the electrical system design are under pre-conceptual stage. The conceptual design for those systems will be developed in the next year. Design requirements for the electrical system set up to develop conceptual design. The same goals as reactor design - enhance safety, reliability, economy, were applied for the development of the requirements. Also the experience of HANARO design and operation was based on. The design requirements for the power distribution, standby power supply, and raceway system will be used for the conceptual design of electrical system.

  18. Return of spent fuel from the Portuguese research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ramalho, A.J.G.; Marques, J.G.; Cardeira, F.M. [Instituto Tecnologico e Nuclear, PO-2686-953 Sacavem (Portugal)

    2000-07-01

    Thirty-nine spent MTR fuel assemblies from the Portuguese Research Reactor were recently returned to the US. Prior to the shipment all assemblies were inspected for corrosion and sipped for determination of fission product leakage. Limitations on the floor loading of the reactor building and on the capacity of the crane prevented the placement and loading of the Transnucleaire IU04 transport cask inside the containment building. The transport cask was thus placed outside, under permanent surveillance, in a support structure built around it. A small transfer cask was used to carry individually the assemblies from the storage racks to the transport cask. A forklift was used as a shuttle between the pool and the IU04. A detailed description of the procedures is given. (author)

  19. Control Rod Reactivity Curves for the Annular Core Research Reactor

    Science.gov (United States)

    Depriest, K. Russell; Kajder, Karen C.; Frye, Jason N.; Denman, Matthew R.

    2009-08-01

    Experiments were conducted at the Annular Core Research Reactor (ACRR) to increase the fidelity of the control rod integral reactivity worth curve. This experiment series was designed to refine the integral reactivity curve used for pulse yield prediction and eliminate the need for operator compensation in the pulse setup. The experiment series consisted of delayed critical and positive period measurements with various ACRR cavity configurations. An improved integral reactivity worth curve for the ACRR control rods has been constructed using the positive period measurements, the delayed critical measurements, and radiation transport modeling of the reactor. A series of prompt period measurements is used to validate that the new control rod curve more accurately predicts the energy yield of the pulse operations. The new reactivity worth curve is compared with the current curve that was developed using traditional approaches.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Romano, A.J.

    1980-01-01

    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.

  2. Testing of Passive Safety System Performance for Higher Power Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    brian G. Woods; Jose Reyes, Jr.; John Woods; John Groome; Richard Wright

    2004-12-31

    This report describes the results of NERI research on the testing of advanced passive safety performance for the Westinghouse AP1000 design. The objectives of this research were: (a) to assess the AP1000 passive safety system core cooling performance under high decay power conditions for a spectrum of breaks located at a variety of locations, (b) to compare advanced thermal hydraulic computer code predictions to the APEX high decay power test data and (c) to develop new passive safety system concepts that could be used for Generation IV higher power reactors.

  3. Conceptual Design of I and C Architecture for a New Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Yong Suk; Park, Jae Kwan; Kim, Taek Kyu; Bae, Sang Hoon; Baang, Dane; Kim, Young Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    NRR) by 2016 was launched by KAERI in 2012. The purposes of the project are to meet domestic industrial needs of a research reactor and to secure an internationally competitive NRR. For instrumentation and control (I and C) systems of the NRR, it shall also be designed to secure the competition of the NRR. The I and C should account for the competition in terms of three aspects: safety, performance and cost. A I and C conceptual design activity should be carried out at the early stage of the project to figure out the I and C on the high level. Constructing I and C architecture is to accomplish the high level design. During the I and C architecture construction, the three aspects should be taken into account. This paper conceptually constructs the I and C architecture for the NRR by comparing to the cases of a Jordan training and research reactor (JRTR) project and a RA-10 multipurpose research reactor. The JRTR is an on-going project launched by KAERI and DAEWOO consortium in 2011. The I and C of the JRTR was digitalized based on the I and C functions of the HANARO research reactor, as shown in Fig. 1. The RA-10 was launched by INVAP of Argentina in 2011. The digital I and C developed by INVAP was built in an OPAL of Australia, as shown in Fig. 2

  4. Hybrid Adsorption-Membrane Biological Reactors for Improved Performance and Reliability of Perchlorate Removal Processes

    Science.gov (United States)

    2008-12-01

    carbon supply for the autotrophic perchlorate reducing bacteria. The membrane used in the reactor is a hollow-fiber microfiltration membrane made from...1 HYBRID ADSORPTION- MEMBRANE BIOLOGICAL REACTORS FOR IMPROVED PERFORMANCE AND RELIABILITY OF PERCHLORATE REMOVAL PROCESSES L.C. Schideman...Center Champaign, IL 61826, USA ABSTRACT This study introduces the novel HAMBgR process (Hybrid Adsorption Membrane Biological Reactor) and

  5. PERFORMANCE IMPROVEMENT OF A CHEMICAL REACTOR BY NONLINEAR NATURAL OSCILLATIONS

    NARCIS (Netherlands)

    RAY, AK

    1995-01-01

    The dynamic behaviour of two coupled continuous stirred tank reactors in sequence is studied when the first reactor is being operated under limit cycle regimes producing self-sustained natural oscillations. The periodic output from the first reactor is then used as a forced input into the second rea

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

    Energy Technology Data Exchange (ETDEWEB)

    Bonnetain, Xavier [AREVA TA, Paris (France)

    2013-07-01

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

  7. Radioisotope Production Plan and Strategy of Kijang Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kye Hong; Lee, Jun Sig [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    This reactor will be located at Kijang, Busan, Korea and be dedicated to produce mainly medical radioisotopes. Tc-99m is very important isotope for diagnosis and more than 80% of radiation diagnostic procedures in nuclear medicine depend on this isotope. There were, however, several times of insecure production of Mo-99 due to the shutdown of major production reactors worldwide. OECD/NEA is leading member countries to resolve the shortage of this isotope and trying to secure the international market of Mo-99. The radioisotope plan and strategy of Kijang Research Reactor (KJRR) should be carefully established to fit not only the domestic but also international demand on Mo-99. The implementation strategy of 6 principles of HLG-MR should be established that is appropriate to national environments. Ministry of Science, ICT and Future Planning and Ministry of Health and welfare should cooperate well to organize the national radioisotope supply structure, to set up the reasonable and competitive pricing of radioisotopes, and to cope with the international supply strategy.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-12-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  11. Fuel burnup calculation of a research reactor plate element

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: nadiasam@gmail.com, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This work consists in simulating the burnup of two different plate type fuel elements, where one is the benchmark MTR of the IAEA, which is made of an alloy of uranium and aluminum, while the other belonging to a typical multipurpose reactor is composed of an alloy of uranium and silicon. The simulation is performed using the WIMSD-5B computer code, which makes use of deterministic methods for solving neutron transport. In developing this task, fuel element equivalent cells were calculated representing each of the reactors to obtain the initial concentrations of each isotope constituent element of the fuel cell and the thicknesses corresponding to each region of the cell, since this information is part of the input data. The compared values of the k∞ showed a similar behavior for the case of the MTR calculated with the WIMSD-5B and EPRI-CELL codes. Relating the graphs of the concentrations in the burnup of both reactors, there are aspects very similar to each isotope selected. The application WIMSD-5B code to calculate isotopic concentrations and burnup of the fuel element, proved to be satisfactory for the fulfillment of the objective of this work. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-03-15

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

  14. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

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

  15. UCLA research reactor relicensing, or guilty until proven innocent

    Energy Technology Data Exchange (ETDEWEB)

    Wegst, W.F.

    1985-11-01

    This paper briefly reviews the history and experiences of the University of California, Los Angeles (UCLA) in attempting to relicense its 100-kW Argonaut research reactor. The process of intervention in US Nuclear Regulatory Commission (NRC) licensing hearings is briefly reviewed. The intervention in the UCLA case, by an antinuclear group called the Committee to Bridge the Gap (CBG), is described. The outcome of the entire proceeding is summarized and opinions are presented on the validity and viability of the licensing/intervention process.

  16. Oak Ridge National Laboratory Research Reactor Experimenters' Guide

    Energy Technology Data Exchange (ETDEWEB)

    Cagle, C.D. (comp.)

    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.

  17. Non-destructive control of cladding thickness of fuel elements for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karlov, Y.; Zhukov, Y.; Chashchin, S

    1997-07-01

    The control method of fuel elements for research reactors by means of measuring beta particles back scattering made it possible to perform complete automatic non-destructive control of internal and external claddings at our plant. This control gives high guarantees of the fuel element correspondence to the requirements. The method can be used to control the three-layer items of different geometry, including plates. (author)

  18. A simple setup for neutron tomography at the Portuguese Nuclear Research Reactor

    CERN Document Server

    Pereira, M A Stanojev; Pugliesi, R

    2012-01-01

    A simple setup for neutron radiography and tomography was recently installed at the Portuguese Research Reactor. The objective of this work was to determine the operational characteristics of the installed setup, namely the irradiation time to obtain the best dynamic range for individual images and the spatial resolution. The performance of the equipment was demonstrated by imaging a fragment of a 17th century decorative tile.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

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

  20. Management of historical waste from research reactors: the Dutch experience

    Energy Technology Data Exchange (ETDEWEB)

    Van Heek, Aliki; Metz, Bert; Janssen, Bas; Groothuis, Ron [NRG, Petten (Netherlands)

    2013-07-01

    Most radioactive waste emerges as well-defined waste streams from operating power reactors. The management of this is an on-going practice, based on comprehensive (IAEA) guidelines. A special waste category however consists of the historical waste from research reactors, mostly originating from various experiments in the early years of the nuclear era. Removal of the waste from the research site, often required by law, raises challenges: the waste packages must fulfill the acceptance criteria from the receiving storage site as well as the criteria for nuclear transports. Often the aged waste containers do not fulfill today's requirements anymore, and their contents are not well documented. Therefore removal of historical waste requires advanced characterization, sorting, sustainable repackaging and sometimes conditioning of the waste. This paper describes the Dutch experience of a historical waste removal campaign from the Petten High Flux research reactor. The reactor is still in operation, but Dutch legislation asks for central storage of all radioactive waste at the COVRA site in Vlissingen since the availability of the high- and intermediate-level waste storage facility HABOG in 2004. In order to comply with COVRA's acceptance criteria, the complex and mixed inventory of intermediate and low level waste must be characterized and conditioned, identifying the relevant nuclides and their activities. Sorting and segregation of the waste in a Hot Cell offers the possibility to reduce the environmental footprint of the historical waste, by repackaging it into different classes of intermediate and low level waste. In this way, most of the waste volume can be separated into lower level categories not needing to be stored in the HABOG, but in the less demanding LOG facility for low-level waste instead. The characterization and sorting is done on the basis of a combination of gamma scanning with high energy resolution of the closed waste canister and low

  1. Neutronic and Thermal-hydraulic Modelling of High Performance Light Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seppaelae, Malla [VTT Technical Research Centre of Finland, P.O.Box 1000, FI02044 VTT (Finland)

    2008-07-01

    High Performance Light Water Reactor (HPLWR), which is studied in EU project 'HPLWR2', uses water at supercritical pressures as coolant and moderator to achieve higher core outlet temperature and thus higher efficiency compared to present reactors. At VTT Technical Research Centre of Finland, functionality of the thermal-hydraulics in the coupled reactor dynamics code TRAB3D/ SMABRE was extended to supercritical pressures for the analyses of HPLWR. Input models for neutronics and thermal-hydraulics were made for TRAB3D/ SMABRE according to the latest HPLWR design. A preliminary analysis was performed in which the capability of SMABRE in the transition from supercritical pressures to subcritical pressures was demonstrated. Parameterized two-group cross sections for TRAB3D neutronics were received from Hungarian Academy of Sciences KFKI Atomic Energy Research Institute together with a subroutine for handling them. PSG, a new Monte Carlo transport code developed at VTT, was also used to generate two-group constants for HPLWR and comparisons were made with the KFKI cross sections and MCNP calculations. (author)

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

    Directory of Open Access Journals (Sweden)

    Meyer Franz

    2010-01-01

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

  3. IAEA Assistance in the development of new research reactor projects

    Energy Technology Data Exchange (ETDEWEB)

    Borio di Tigliole, Andrea; Bradley, Ed; Zhukova, Anastasia; Adelfang, Pablo [International Atomic Energy Agency, Research Reactor Section, Vienna (Austria); Shokr, Amgad [International Atomic Energy Agency, Research Reactor Safety Section, Vienna (Austria); Ridikas, Danas [International Atomic Energy Agency, Physics Section, Vienna (Austria)

    2015-08-15

    A research reactor (RR) project is a major undertaking that requires careful preparation, planning, implementation and investment in time, money, and human resources. In recent years, the interest of IAEA Member States in developing RR programmes has grown significantly, and currently, several Member States are in different stages of new RR projects. The majority of these countries are building their first RR as a key national facility for the development of their nuclear science and technology programmes, including nuclear power. In order to support Member States in such efforts, the IAEA in 2012 published the Nuclear Energy Series Report No. NP-T-5.1 on Specific Considerations and Milestones for a Research Reactor Project. To provide further support, the IAEA also published a document to assist Member States in the preparation of the bid invitation specification for the purchase of a RR. The IAEA will also continue to provide assistance for human resources development of the Member States establishing their first RR, and to facilitate sharing experience and knowledge among Member States through its programmatic activities including expert mission services, technical meetings, training courses and workshops addressing relevant technical and safety topics. This paper presents the IAEA assistance and services provided to the Member States considering new RRs, with particular emphasis on those establishing their first RR, including elaboration on the services mentioned above.

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

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbalm, K.F. [comp.

    1995-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-15

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

  6. Research and development studies on the seismic behaviour of the PEC fast reactor (safety analysis detailed report no. 8)

    Energy Technology Data Exchange (ETDEWEB)

    Martelli, A.; Forni, M.; Masoni, P.; Maresca, G.; Castoldi, A.; Muzzi, F. (ENEA, Rome (Italy); Ansaldo Spa, Genoa (Italy); ISMES Spa, Bergamo (Italy))

    1988-01-15

    This paper presents the main features and results of the numerical and experimental studies that were carried out by ENEA (Italian Commission for Alternative Energy Sources) for the seismic verification of the Italian PEC fast reactor test facility. More precisely, the paper focuses on the wide-ranging research and development programme that has been performed (and recently completed) on the reactor building, the reactor-block, the main vessel, the core and the shutdown system. The needs of these detailed studies are stressed and the feed-backs on the design, necessary safisfy the seismic safety requirements, are recalled. The general validity of the analyses in the framework of the research and development activities for nuclear reactor is also pointed out.

  7. Reprocessing of research reactor fuel the Dounreay option

    Energy Technology Data Exchange (ETDEWEB)

    Cartwright, P.

    1997-08-01

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

  8. System Requirements Analysis for a Computer-based Procedure in a Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaek Wan; Jang, Gwi Sook; Seo, Sang Moon; Shin, Sung Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    This can address many of the routine problems related to human error in the use of conventional, hard-copy operating procedures. An operating supporting system is also required in a research reactor. A well-made CBP can address the staffing issues of a research reactor and reduce the human errors by minimizing the operator's routine tasks. A CBP for a research reactor has not been proposed yet. Also, CBPs developed for nuclear power plants have powerful and various technical functions to cover complicated plant operation situations. However, many of the functions may not be required for a research reactor. Thus, it is not reasonable to apply the CBP to a research reactor directly. Also, customizing of the CBP is not cost-effective. Therefore, a compact CBP should be developed for a research reactor. This paper introduces high level requirements derived by the system requirements analysis activity as the first stage of system implementation. Operation support tools are under consideration for application to research reactors. In particular, as a full digitalization of the main control room, application of a computer-based procedure system has been required as a part of man-machine interface system because it makes an impact on the operating staffing and human errors of a research reactor. To establish computer-based system requirements for a research reactor, this paper addressed international standards and previous practices on nuclear plants.

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

    Energy Technology Data Exchange (ETDEWEB)

    Rosenthal, Murray Wilford [ORNL

    2009-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-07-01

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

  13. Dosimetry at the Portuguese research reactor using thermoluminescence measurements and Monte Carlo calculations.

    Science.gov (United States)

    Fernandes, A C; Gonçalves, I C; Santos, J; Cardoso, J; Santos, L; Ferro Carvalho, A; Marques, J G; Kling, A; Ramalho, A J G; Osvay, M

    2006-01-01

    This work presents an extensive study on Monte Carlo radiation transport simulation and thermoluminescent (TL) dosimetry for characterising mixed radiation fields (neutrons and photons) occurring in nuclear reactors. The feasibility of these methods is investigated for radiation fields at various locations of the Portuguese Research Reactor (RPI). The performance of the approaches developed in this work is compared with dosimetric techniques already existing at RPI. The Monte Carlo MCNP-4C code was used for a detailed modelling of the reactor core, the fast neutron beam and the thermal column of RPI. Simulations using these models allow to reproduce the energy and spatial distributions of the neutron field very well (agreement better than 80%). In the case of the photon field, the agreement improves with decreasing intensity of the component related to fission and activation products. (7)LiF:Mg,Ti, (7)LiF:Mg,Cu,P and Al(2)O(3):Mg,Y TL detectors (TLDs) with low neutron sensitivity are able to determine photon dose and dose profiles with high spatial resolution. On the other hand, (nat)LiF:Mg,Ti TLDs with increased neutron sensitivity show a remarkable loss of sensitivity and a high supralinearity in high-intensity fields hampering their application at nuclear reactors.

  14. Thermal hydraulic analysis of reactivity accidents in MTR research reactors using RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    El-Sahlamy, N.; Khedr, A. [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt); D' Auria, F.D. [Pisa Univ. (Italy). Facolta di Ingegneria

    2015-12-15

    The present paper comes in the line with the international approach which use the best estimate codes, instead of conservative codes, to get more realistic prediction of system behavior under off-normal reactor conditions. The aim of the current work is to apply this approach using the thermal-hydraulic system code RELAP5/Mod3.3 in a reassessment of safety of the IAEA benchmark 10 MW Research Reactor. The assessment is performed for both slow and fast reactivity insertion transients at initial power of 1.0 W. The reactor power is calculated using the RELA5 point kinetic model. The reactivity feedback terms are considered in two steps. In the first step the feedback from changes in water density and fuel temperature (Doppler effects) are considered. In the second step the feedback from the water temperature changes is added. The results from the first step are compared with that published in IAEA-TECDOC-643 benchmarks. The comparison shows that RELAP5 over predicts the peak power and consequently the fuel, clad and coolant temperatures in case of fast reactivity insertion. The results from the second step show unjustified values for reactor power. Therefore, the model of reactivity feedback from water temperature changes in the RELAP5 code may have to be reviewed.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hakan Ozaltun & Herman Shen

    2011-11-01

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

  16. Updating of PGAA system at HANARO research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeo, H. J.; Kim, S. H.; Moon, J. H.; Jeong, Y. S.; Kim, Y. J. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    In this study, updating of Prompt Gamma-ray neutron Activation Analysis system (PGAA) has been carried out to obtain the best, optimal condition through the improvement of neutron flux and reduction of background of PGAA facility which is installed on the ST 1 horizontal beam port at HANARO research reactor. Both diffracted beam profiling's conditions and the neutron diffraction of pyrolytic graphite crystals are investigated by BF{sub 3} counter, laser and optical diffraction angle control method to confirm the beam convergence rate. Also, the effects of interference materials such as aluminum sample holder, teflon holder and Teflon wire appeared from analyzing elemental constituent are investigated with single - and Compton mode. After readjusting of system, the neutron flux measured was 8.1{+-}0.2 x 10{sup 7} n{center_dot}cm{sup -2}{center_dot}s{sup -1} increasing about 40%, to be expected the improved analytical sensitivity.

  17. Radiopharmaceuticals developed at the University of Missouri research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ketring, A.R.; Ehrhardt, G.J. [Univ. of Missouri, Columbia, MO (United States); Day, D.E. [Univ. of Missouri, Rolla, MO (United States)

    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. The neutron texture diffractometer at the China Advanced Research Reactor

    Science.gov (United States)

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

    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)

  19. Computational simulation of fuel burnup estimation for research reactors plate type

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos, E-mail: nadiasam@gmail.com [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The aim of this study is to estimate the spatial fuel burnup, through computational simulation, in two research reactors plate type, loaded with dispersion fuel: the benchmark Material Test Research - International Atomic Energy Agency (MTR-IAEA) and a typical multipurpose reactor (MR). The first composed of plates with uranium oxide dispersed in aluminum (UAlx-Al) and a second composed with uranium silicide (U{sub 3}Si{sub 2}) dispersed in aluminum. To develop this work we used the deterministic code, WIMSD-5B, which performs the cell calculation solving the neutron transport equation, and the DF3DQ code, written in FORTRAN, which solves the three-dimensional neutron diffusion equation using the finite difference method. The methodology used was adequate to estimate the spatial fuel burnup , as the results was in accordance with chosen benchmark, given satisfactorily to the proposal presented in this work, even showing the possibility to be applied to other research reactors. For future work are suggested simulations with other WIMS libraries, other settings core and fuel types. Comparisons the WIMSD-5B results with programs often employed in fuel burnup calculations and also others commercial programs, are suggested too. Another proposal is to estimate the fuel burnup, taking into account the thermohydraulics parameters and the Xenon production. (author)

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

    Science.gov (United States)

    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.

  1. Scalable synthesis of ionic liquids: comparison of performances of microstructured and stirred batch reactors

    OpenAIRE

    Iken, Hicham; Guillen, Frédéric; Chaumat, Hélène; Mazières, Marie-Rose; Plaquevent, Jean-Christophe; Tzedakis, Théodore

    2012-01-01

    A range of alkylpyridinium bromide ionic liquids have been synthesized in a stirred reactor at multigram scale and characterized by physical methods (viscosity, conductivity, melting point, electrochemical window, and water content). One ionic liquid, octylpyridinium bromide, was chosen to be synthesized in both macro and reduced scale reactors, in order to compare its performance and to afford evidence of the advantages of a cross channel micro reactor (channel width = 1 mm) compared to a st...

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

    Energy Technology Data Exchange (ETDEWEB)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

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

  3. An immobilized cell reactor with simultaneous product separation. II. Experimental reactor performance.

    Science.gov (United States)

    Dale, M C; Okos, M R; Wankat, P C

    1985-07-01

    The simultaneous separation of volatile fermentation products from product-inhibited fermentations can greatly increase the productivity of a bioreactor by reducing the product concentration in the bioreactor, as well as concentrating the product in an output stream free of cells, substrate, or other feed impurities. The Immobilized Cell Reactor-Separator (ICRS) consists of two column reactors: a cocurrent gas-liquid "enricher" followed by a countercurrent "stripper" The columns are four-phase tubular reactors consisting of (1) an inert gas phase, (2) the liquid fermentation broth, (3) the solid column internal packing, and (4) the immobilized biological catalyst or cells. The application of the ICRS to the ethanol-from-whey-lactose fermentation system has been investigated. Operation in the liquid continuous or bubble flow regime allows a high liquid holdup in the reactor and consequent long and controllable liquid residence time but results in a high gas phase pressure drop over the length of the reactor and low gas flow rates. Operation in the gas continuous regime gives high gas flow rates and low pressure drop but also results in short liquid residence time and incomplete column wetting at low liquid loading rates using conventional gas-liquid column packings. Using cells absorbed to conventional ceramic column packing (0.25-in. Intalox saddles), it was found that a good reaction could be obtained in the liquid continuous mode, but little separation, while in the gas continuous mode there was little reaction but good separation. Using cells sorbed to an absorbant matrix allowed operation in the gas continuous regime with a liquid holdup of up to 30% of the total reactor volume. Good reaction rates and product separation were obtained using this matrix. High reaction rates were obtained due to high density cell loading in the reactor. A dry cell density of up to 92 g/L reactor was obtained in the enricher. The enricher ethanol productivity ranged from 50 to 160

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

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, P.; Zhang, Hongxia; Bao, W. [Department of Physics, Beijing Key Laboratory of Opto-electronic Functional Materials & Micro-nano Devices, Renmin University of China, Beijing 100872 (China); Schneidewind, A. [Jülich Center for Neutron Science (JCNS), Forschungszentrum Jülich GmbH, Outstation at Heinz MaierCLeibnitz Zentrum (MLZ), D-85747 Garching (Germany); Link, P. [Heinz Maier-Leibnitz Zentrum, Technische Universität München, D-85748 Garching (Germany); Grünwald, A.T.D. [II. Physikalisches Institut, Universität zu Köln, D-50937 Köln (Germany); Georgii, R. [Heinz Maier-Leibnitz Zentrum, Technische Universität München, D-85748 Garching (Germany); Hao, L.J.; Liu, Y.T. [China Institute of Atomic Energy, PO Box-275-30, Beijing 102413 (China)

    2016-06-11

    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 ×10{sup 7} n/cm{sup 2}/s at neutron incident energy E{sub i}=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.

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

    Science.gov (United States)

    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.

  6. Research reactor power controller design using an output feedback nonlinear receding horizon control method

    Energy Technology Data Exchange (ETDEWEB)

    Etchepareborda, Andres [Department of Nuclear Engineering, Argentine National Atomic Energy Commission, Centro Atomico Bariloche, Av. E. Bustillo 9500, Bariloche 8400 (Argentina)]. E-mail: etche@cab.cnea.gov.ar; Lolich, Jose [INVAP S.E., Moreno 1089, Bariloche 8400 (Argentina)

    2007-02-15

    A constrained, output feedback nonlinear receding horizon control (NRHC) method is applied to design a research reactor power controller. The method uses a nonlinear plant model subject to state, control and terminal set constraints; a nonlinear cost function; and a high gain observer. The controller regulates reactor power from 1% to 100% of full power; considers known disturbances, such as reactivity insertions and changes in core inlet flow and temperature; and includes upper limits constraints on neutron flux, neutron flux rate, core outlet temperature and core inlet-outlet temperature difference. Simulation results show an excellent performance for power regulation and known disturbances rejection: all process variables are kept within the admissible limits avoiding the actuation of the safety systems.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-01

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

  8. 75 FR 57080 - In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order...

    Science.gov (United States)

    2010-09-17

    ... COMMISSION In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order... Aerotest Operations, Inc., (Aerotest, the licensee) is the holder of Facility Operating License No. R-98 which authorizes the possession, use, and operation of the Aerotest Radiography and Research Reactor...

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

    Science.gov (United States)

    2010-07-13

    ... COMMISSION In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order Approving Indirect Transfer of Facility Operating License and Conforming Amendment I. Aerotest Operations..., use and operation of the Aerotest Radiography and Research Reactor (ARRR) located in San Ramon...

  10. 78 FR 46618 - Order Prohibiting Operation of Aerotest Radiography and Research Reactor

    Science.gov (United States)

    2013-08-01

    ... COMMISSION Order Prohibiting Operation of Aerotest Radiography and Research Reactor I. Aerotest Operations... Licensing of Production and Utilization Facilities.'' The license authorizes the operation of the Aerotest Radiography and Research Reactor (ARRR) in accordance with the conditions specified therein. The ARRR is...

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  14. Main technical options of the Jules Horowitz reactor project to achieve high flux performances and high safety level

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, 13 - Saint-Paul-lez-Durance (France)]|[CEA Saclay Dir. de l' Energie Nucleaire DEN, 91 - Gif sur Yvette (France)

    2003-07-01

    Since the shutdown of the SILOE reactor in 1997, the OSIRIS reactor has ensured the needs regarding technological irradiation at CEA including those of its industrial partners and customers. The Jules Horowitz Reactor will replace it and will offer a quite larger experimental field. It has the ambition to provide the necessary nuclear data and to maintain a fission research capability in Europe after 2010. The Jules Horowitz Reactor will represent a significant step in terms of performances and experimental capabilities. This paper will present the main design option resulting from the preliminary studies. The choice of the specific power around 600 kW/I for the reference core configuration is a key decision to ensure the required flux level. Consequently many choices have to be made regarding the materials used in the core and the fuel element design. These involve many specific qualifications including codes validation. The main safety options are based on: - A safety approach based upon the defence-in-depth principle. - A strategy of generic approaches to assess experimental risks in the facility. - Internal events analysis taking into account risks linked to reactor and experiments (e.g., radioactive source-term). - Systematic consideration of external hazards (e.g., earthquake, airplane crash) and internal hazards. - Design of containment to manage and mitigate a severe reactor accident (consideration of 'BORAX' accident, according to french safety practice for MTRs, beyond design basis reactivity insertion accident, involving core melting and core destruction phenomena). (authors)

  15. Recent performance experience with US light water reactor self-actuating safety and relief valves

    Energy Technology Data Exchange (ETDEWEB)

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  16. Delayed Gamma Measurements in Different Nuclear Research Reactors Bringing Out the Importance of the Delayed Contribution in Gamma Flux Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Fourmentel, D.; Radulovic, V.; Barbot, L.; Villard, J-F. [Alternative Energies and Atomic Energy Commission, CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, 13108 Saint- Paul-Lez-Durance (France); Zerovnik, G.; Snoj, L. [Reactor Physics Department, Jozef Stefan Institute, SI-1000 Ljubljana (Slovenia); Tarchalski, M.; Pytel, K. [National Centre for Nuclear Research A. Soltana 7, 05-400 Swierk (Poland); Malouch, F. [Alternative Energies and Atomic Energy Commission - CEA, DEN, DM2S, Saclay, 91191, Gif-sur-Yvette (France)

    2015-07-01

    Neutron and gamma flux levels are key parameters in nuclear research reactors. In Material Testing Reactors, such as the future Jules Horowitz Reactor, under construction at the French Alternative Energies and Atomic Energy Commission (CEA Cadarache, France), the expected gamma flux levels are very high (nuclear heating is of the order of 20 W/g at 100 MWth). As gamma rays deposit their energy in the reactor structures and structural materials it is important to take them into account when designing irradiation devices. There are only a few sensors which allow measurements of the nuclear heating ; a recent development at the CEA Cadarache allows measurements of the gamma flux using a miniature ionization chamber (MIC). The measured MIC response is often compared with calculation using modern Monte Carlo (MC) neutron and photon transport codes, such as TRIPOLI-4 and MCNP6. In these calculations only the production of prompt gamma rays in the reactor is usually modelled thus neglecting the delayed gamma rays. Hence calculations and measurements are usually in better accordance for the neutron flux than for the gamma flux. In this paper we study the contribution of delayed gamma rays to the total MIC signal in order to estimate the systematic error in gamma flux MC calculations. In order to experimentally determine the delayed gamma flux contributions to the MIC response, we performed gamma flux measurements with CEA developed MIC at three different research reactors: the OSIRIS reactor (MTR - 70 MWth at CEA Saclay, France), the TRIGA MARK II reactor (TRIGA - 250 kWth at the Jozef Stefan Institute, Slovenia) and the MARIA reactor (MTR - 30 MWth at the National Center for Nuclear Research, Poland). In order to experimentally assess the delayed gamma flux contribution to the total gamma flux, several reactor shut down (scram) experiments were performed specifically for the purpose of the measurements. Results show that on average about 30 % of the MIC signal is due to

  17. Advanced reactor material research requirements; Necesidades de investigacion en materiales para reactores avanzados

    Energy Technology Data Exchange (ETDEWEB)

    Greene, C. A.; Muscara, J.; Srinivasan, M.

    2003-07-01

    The metal and graphite components used in high temperature gas-cooled reactors (HTGR) may suffer physical-chemical alterations, irradiation damage and mechanical alterations. Their failure may call the security of these reactors into question by affecting the integrity of the pressure control system, core geometry or its cooling, among other aspects. This article analyses the work currently being done in the matter by the US Nuclear Regulatory Commission. (Author)

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

    Science.gov (United States)

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

    1976-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-11-03

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

  20. Improved performance of parallel surface/packed-bed discharge reactor for indoor VOCs decomposition: optimization of the reactor structure

    Science.gov (United States)

    Jiang, Nan; Hui, Chun-Xue; Li, Jie; Lu, Na; Shang, Ke-Feng; Wu, Yan; Mizuno, Akira

    2015-10-01

    The purpose of this paper is to develop a high-efficiency air-cleaning system for volatile organic compounds (VOCs) existing in the workshop of a chemical factory. A novel parallel surface/packed-bed discharge (PSPBD) reactor, which utilized a combination of surface discharge (SD) plasma with packed-bed discharge (PBD) plasma, was designed and employed for VOCs removal in a closed vessel. In order to optimize the structure of the PSPBD reactor, the discharge characteristic, benzene removal efficiency, and energy yield were compared for different discharge lengths, quartz tube diameters, shapes of external high-voltage electrode, packed-bed discharge gaps, and packing pellet sizes, respectively. In the circulation test, 52.8% of benzene was removed and the energy yield achieved 0.79 mg kJ-1 after a 210 min discharge treatment in the PSPBD reactor, which was 10.3% and 0.18 mg kJ-1 higher, respectively, than in the SD reactor, 21.8% and 0.34 mg kJ-1 higher, respectively, than in the PBD reactor at 53 J l-1. The improved performance in benzene removal and energy yield can be attributed to the plasma chemistry effect of the sequential processing in the PSPBD reactor. The VOCs mineralization and organic intermediates generated during discharge treatment were followed by CO x selectivity and FT-IR analyses. The experimental results indicate that the PSPBD plasma process is an effective and energy-efficient approach for VOCs removal in an indoor environment.

  1. Effect of Sulfur Source on the Performance and Metal Retention of Methanol-Fed UASB Reactors

    NARCIS (Netherlands)

    Zandvoort, M.H.; Hullebusch, van E.D.; Gieteling, J.; Lettinga, G.; Lens, P.N.L.

    2005-01-01

    The effect of a sulfur source on the performance and metal retention of methanol-fed upflow anaerobic sludge bed (UASB) reactors was investigated. For this purpose, two UASB reactors were operated with cobalt preloaded granular sludge (1 mM CoCl2; 30° C; 24 h) at an organic loading rate (OLR) of 5 g

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-09-01

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

  3. Atmospheric dispersion of argon-41 from anuclear research reactor: measurement and modeling of plume geometry and gamma radiation field

    DEFF Research Database (Denmark)

    Lauritzen, Bent; Astrup, Poul; Drews, Martin

    2003-01-01

    An atmospheric dispersion experiment was conducted using a visible tracer along with the routine release of argon-41 from the BR1 research reactor in Mol, Belgium. Simultaneous measurements of plume geometry and radiation fields for argon-41 decay were performed as well as measurements of the argon...

  4. Atmospheric dispersion of argon-41 from anuclear research reactor: measurement and modeling of plume geometry and gamma radiation field

    DEFF Research Database (Denmark)

    Lauritzen, Bent; Astrup, Poul; Drews, Martin

    2003-01-01

    An atmospheric dispersion experiment was conducted using a visible tracer along with the routine release of argon-41 from the BR1 research reactor in Mol, Belgium. Simultaneous measurements of plume geometry and radiation fields for argon-41 decay were performed as well as measurements of the argon...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  6. Progress of Research on Demonstration Fast Reactor Main Pipe Material

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    The main characteristics of the sodium pipe system in demonstration fast reactor are high-temperature, thin-wall and big-caliber, which is different from the high-pressure and thick-wall of the pressurized water reactor system, and the system is long-term

  7. Researches on a reactor core in heavy ion inertial fusion

    CERN Document Server

    Kondo, S; Iinuma, T; Kubo, K; Kato, H; Kawata, S; Ogoyski, A I

    2016-01-01

    In this paper a study on a fusion reactor core is presented in heavy ion inertial fusion (HIF), including the heavy ion beam (HIB) transport in a fusion reactor, a HIB interaction with a background gas, reactor cavity gas dynamics, the reactor gas backflow to the beam lines, and a HIB fusion reactor design. The HIB has remarkable preferable features to release the fusion energy in inertial fusion: in particle accelerators HIBs are generated with a high driver efficiency of ~30-40%, and the HIB ions deposit their energy inside of materials. Therefore, a requirement for the fusion target energy gain is relatively low, that would be ~50 to operate a HIF fusion reactor with a standard energy output of 1GW of electricity. In a fusion reactor the HIB charge neutralization is needed for a ballistic HIB transport. Multiple mechanical shutters would be installed at each HIB port at the reactor wall to stop the blast waves and the chamber gas backflow, so that the accelerator final elements would be protected from the ...

  8. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-10-03

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

  10. Reactor

    Science.gov (United States)

    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.

  11. The application of research reactor Maria for analysis of thorium use in nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Chwaszczewski, S.; Andrzejewski, K.; Myslek-Laurikainen, B.; Pytel, B.; Szczurek, J. [Dep. Thorium Project, Institute of Atomic Energy POLATOM, 05-400 Otwock-Swierk (Poland); Polkowska-Motrenko, H. [Institute of Nuclear Chemistry and Technology, ul.Dorodna 16 03-195 Warszawa (Poland)

    2010-07-01

    The MARIA reactor, pool-type light-water cooled and beryllium moderated nuclear research reactor was used to evaluate the {sup 233}U breeding during the experimental irradiation of the thorium samples. The level of impurities concentrations was determined using ICP-MS method. The associated development of computer programs for analysis of application of thorium in EPR reactor consist of PC version of CORD-2/GNOMER system are presented. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2008-04-01

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

  15. Severe Accidents and New Reactors. Twenty Years of Research; Accidents severos y nuevos reactores. Veinte anos de investigacion

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J.

    2008-07-01

    A review was done on the main activities performed by the Programme for Nuclear Safety of CIEMAT in the field of nuclear reactor safety from 1985 to 2005. It covers the areas of severe accident and source term, advanced and passive reactors, containments analyses and plant applications. It is emphasized CIEMATs participation in national and international projects mainly in those supported by CSN, OECD and the EU. At the same time, experimental and analytical capabilities set up at CIEMAT, as PECA, RECA and GIRS for simulating aerosol pool scrubbing phenomena, hydrogen catalytic recombiner and sprays are been presented, together with an Annex on Generation IV. Two chapters were added, one on the nuclear power reactors in the world and another about the safety systems and principles. (Author)

  16. Effect of auxotrophies on yeast performance in aerated fed-batch reactor

    Energy Technology Data Exchange (ETDEWEB)

    Landi, Carmine; Paciello, Lucia [Dept. Ingegneria Industriale, Universita di Salerno, Via Ponte Don Melillo, 84084 Fisciano, Salerno (Italy); Alteriis, Elisabetta de [Dept. Biologia Strutturale e Funzionale, Universita degli Studi di Napoli ' Federico II' , Via Cinthia, 80100 Napoli (Italy); Brambilla, Luca [Dept. Biotecnologie e Bioscienze, Universita Milano-Bicocca, Piazza della Scienza, 20126 Milano (Italy); Parascandola, Palma, E-mail: pparascandola@unisa.it [Dept. Ingegneria Industriale, Universita di Salerno, Via Ponte Don Melillo, 84084 Fisciano, Salerno (Italy)

    2011-10-28

    Highlights: Black-Right-Pointing-Pointer The paper contributes to fill the gap existing between the basic and applied research. Black-Right-Pointing-Pointer Mathematical model sheds light on the physiology of auxotrophic yeast strains. Black-Right-Pointing-Pointer Yeast behavior in fed-batch is influenced by biological and environmental determinants. Black-Right-Pointing-Pointer Process optimization would make possible the production of heterologous proteins which are not yet on the market. -- Abstract: A systematic investigation on the effects of auxotrophies on the performance of yeast in aerated fed-batch reactor was carried out. Six isogenic strains from the CEN.PK family of Saccharomyces cerevisiae, one prototroph and five auxotrophs, were grown in aerated fed-batch reactor using the same operative conditions and a proper nutritional supplementation. The performance of the strains, in terms of final biomass decreased with increasing the number of auxotrophies. Auxotrophy for leucine exerted a profound negative effect on the performance of the strains. Accumulation of reactive oxygen species (ROS) in the cells of the strain carrying four auxotrophies and its significant viability loss, were indicative of an oxidative stress response induced by exposure of cells to the environmental conditions. The mathematical model was fundamental to highlight how the carbon flux, depending on the number and type of auxotrophies, was diverted towards the production of increasingly large quantities of energy for maintenance.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-02-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, Young Jun

    2008-11-15

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

  19. Research on the usage of a deep sea fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Otsubo, Akira; Kowata, Yasuki [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-09-01

    Many new types of fast reactors have been studied in PNC. A deep sea fast reactor has the highest realization probability of the reactors studied because its development is desired by many specialists of oceanography, meteorology, deep sea bottom oil field, seismology and so on and because the development does not cost big budget and few technical problems remain to be solved. This report explains the outline and the usage of the reactor of 40 kWe and 200 to 400 kWe. The reactor can be used as a power source at an unmanned base for long term climate prediction and the earth science and an oil production base in a deep sea region. On the other hand, it is used for heat and electric power supply to a laboratory in the polar region. In future, it will be used in the space. At the present time, a large FBR development plan does not proceed successfully and a realization goal time of FBR has gone later and later. We think that it is the most important to develop the reactor as fast as possible and to plant a fast reactor technique in our present society. (author)

  20. Jordan's First Research Reactor Project: Driving Forces, Present Status and the Way Ahead

    Energy Technology Data Exchange (ETDEWEB)

    Xoubi, Ned, E-mail: Ned@Xoubi.co [Jordan Atomic Energy Commission (JAEC), P.O.Box 70, Shafa Badran, 11934 Amman (Jordan)

    2011-07-01

    In a gigantic step towards establishing Jordan's nuclear power program, Jordan Atomic Energy Commission (JAEC) is building the first nuclear research and test reactor in the Kingdom. The new reactor will serve as the focal point for Jordan Center for Nuclear Research (JCNR), a comprehensive state of the art nuclear center not only for Jordan but for the whole region, the center will include in addition to the reactor a radioisotopes production plant, a nuclear fuel fabrication plant, a cold neutron source (CNS), a radioactive waste treatment facility, and education and training center. The JRTR reactor is the only research reactor new build worldwide in 2010, it is a 5 MW light water open pool multipurpose reactor, The reactor core is composed of 18 fuel assemblies, MTR plate type, with 19.75% enriched uranium silicide (U{sub 3}Si{sub 2}) in an aluminum matrix. It is reflected on all sides by beryllium and graphite blocks. Reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45x10{sup 14} cm{sup -2}s{sup -1}. The reactor reactivity is controlled by four Hafnium Control Absorber Rods (CAR). Jordan Center for Nuclear Research is located in Ramtha city, it is owned by Jordan Atomic Energy Commission (JAEC), and is contracted to Korea Atomic Energy Research Institute (KAERI) and Daewoo E and C. The JCNR project is a 56 months EPC fixed price contract for the design engineering, construction, and commissioning the JCNR reactor, and other nuclear facilities. The project presents many challenges for both the owner and the contractor, being the first nuclear reactor for Jordan, and the first nuclear export for Korea. The driving forces, present status and the way ahead will be presented in this paper. (author)

  1. Research Performance Progress Report

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hye -Sook [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-09-10

    The major goals of this project is to develop a suite of diagnostics to probe magnetic fields generated by the dynamics of high velocity interpenetrating plasma flows relevant to astrophysical collisionless shocks. Collisionless shocks are common in the universe and are responsible for decelerating and thermalizing supersonic plasma flows and accelerating a fraction of the incident particles to high energies. When high velocity, low density, plasma flows interact in astrophysics, turbulent electrostatic and electromagnetic waves are generated due to plasma instabilities, such as the Weibel instability. This can lead to localized pockets of very strong magnetic field generation. The net result is that the plasmas stagnate in what is called a collisionless shock. Understanding these enigmatic interactions requires well-controlled laboratory experiments able to validate the theory and the simulations. Time and spatially resolved magnetic field diagnostics are key to probing these frontier plasma dynamics, relevant to both astrophysics and laboratory applications of plasma physics. This project will enable us to develop the necessary diagnostics for this experiment on NIF. Our team has vast experience in performing laser experiments, theory, simulations and diagnostic development and is ideally suited for carrying out this work.

  2. New research heavy-water reactors with the capacities 25 MWt and 100 MWt for the production of radionuclides with high specific radioactivity

    Energy Technology Data Exchange (ETDEWEB)

    Kiselev, G.V.; Myrtsimova, L.A.; Shvedov, I.I. [Institute of Theroretical and Experimental Physics, Moscow (Russian Federation). State Scientific Center of the Russian Federation

    1997-10-01

    Information about the design and performance of two new heavy-water research reactors, HWR-25 and HWR-1 is given. These reactors have multiple-functions: neutron investigations, experimental grounding of fuel for the power reactors, production of radionuclides and so on. Reactor HWR-25 has the capacity 25 MWt and maximal thermal neutron density flux in the reflector is 4,6.1024 cm{sup 2}s{sup -1}. Burn-up of fuel is 50%, duration of the campaign is 105 days. Reactor HWR-100 has the capacity 100 MWt and maximal thermal neutron density flux in the reflector is approximately 1,2.1015 cm{sup 2}s{sup -1}. Both reactors have a large amount of channels for the place of target for the production of radionuclides 2 refs., 6 tabs., 3 figs.

  3. Research on Design and Intelligent Control and Related Stirring Performance Parameters based on Reactor%基于反应釜智能控制设计及相关搅拌性能参数的研究

    Institute of Scientific and Technical Information of China (English)

    张中领

    2015-01-01

    The reactor has a broad application of a chemical reactor,which has a strong time-delay,nonlinear,time-var-ying and other defects,it is difficult to establish a precise mathematical model for the reaction system ,and the classic , modern control optimal control theory in the reactor control is also not fully reflected.To solve the above problems,this pa-per based on the analysis of the heat transfer characteristics of the reactor,designed a reactor intelligent control system. Through the intelligent controller effective control the reactor temperature gradient.It is great significance of improving the quality of chemical production.%反应釜是一种有着广泛应用的化学反应系统,其具有明显的时滞性、非线性和时变性等缺陷,难以建立精确的数学模型;而经典、现代控制理论的最优控制在反应釜控制中也无法完全体现出来。针对上述问题,在分析反应釜传热特性的基础上,设计了一种反应釜智能控制系统。可以通过控制执行器对反应釜内温度及其梯度进行有效控制,为提高化工生产质量提供了有力保障,具有重要的现实意义。

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

    CERN Document Server

    Körner, S; Von Tobel, P; Rauch, H

    2001-01-01

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

  5. Current status, research progress and future plan of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sardjono, Y.; Syarip; Tjiptono, T.W. [Yogyakarta Nuclear Research Center, Batan (Indonesia)

    1999-10-01

    The current status, research progress and future plan of the Kartini Research Reactor (KRR) is presented. The measurements of axial burn-up distributions for each fuel element by gamma scanning techniques, core axial power distribution display, fuel management for safeguards purpose as well as some research progress activities i.e.; utilization of beamport for: neutron radiography, application neutron activation analysis and history record of KRR power operations is also presented. The KRR is 100 kW pool water reactor type which uses natural circulation and provided by: five beamports in which one of them already coupled with natural uranium subcritical assembly, two thermalizing columns in which one of them is prepared for developing Boron Neutron Capture Therapy (BNCT), two rabbit systems utilized for special analysis uranium ore by delayed neutron counting techniques, one center timbre and 40 irradiation rack (lazy susan) for neutron activation analysis. The KRR was constructed as a second research reactor in Indonesia with special purpose for training and education, high safety margin with involve in high negative temperature coefficient which achieved its first criticality on January 25, 1979. The maximum power level on first criticality is 50 kW and since August 1981 up to now is operating 100 kW. Base on the KRR design limit, it is planned to increase the power level up to 250 kW in the future plan. The preliminary activities such as Non Destructive Testing (NDT) for some reactor components especially water tank and thermal column should be done before decided to increase power level. (author)

  6. Biblical Performance Criticism: Performance as Research

    Directory of Open Access Journals (Sweden)

    David Rhoads

    2010-03-01

    Full Text Available The purpose of this paper is to explore the New Testament as performance literature. Recent scholarship has brought to the fore the importance of interpreting the New Testament in the context of the oral cultures of the first century. This leads us to ask about the oral/performance dynamics of the “writings” now in the New Testament canon. The paper is comprised of three parts. Part 1 offers some reasons why the New Testament writings, if they are to be appropriately interpreted in their original oral context, must be treated as performance literature. Part 2 outlines performance criticism as a methodology for this task—constructing first-century scenarios of performance as a context for interpretation, reorienting and expanding methodological approaches to the New Testament as oral literature, and employing contemporary performance as a research tool. Part 3 represents the bulk of the paper, laying out in detail the idea of performance as research, arguing that what we can learn from contemporary performances may help us interpret the New Testament writings in the context of the oral cultures of the first century.

  7. Numerical Simulation of Flow Field in Flow-guide Tank of China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The flow-guide tank of China advanced research reactor (CARR) is located at the top of the reactor vessel and connected with the inlet coolant pipe. It acts as a reactor inlet coolant distributor and plays an important role in reducing the flow-induced vibration of the internal components of the reactor core. Several designs of the flow-guide tank have been proposed, however, the final design option has to be made after detailed investigation of the velocity profile within the flow-guide tank for each configuration.

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

    Directory of Open Access Journals (Sweden)

    M.H. Altaf

    2014-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

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

    NARCIS (Netherlands)

    Wassink, J.

    2013-01-01

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

  11. Highest average burnups achieved by MTR fuel elements of the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Damy, Margaret A.; Terremoto, Luis A.A.; Silva, Jose E.R.; Silva, Antonio Teixeira e; Castanheira, Myrthes; Teodoro, Celso A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear (CEN)]. E-mail: madamy@ipen.br

    2007-07-01

    Different nuclear fuels were employed in the manufacture of plate type at IPEN , usually designated as Material Testing Reactor (MTR) fuel elements. These fuel elements were used at the IEA-R1 research reactor. This work describes the main characteristics of these nuclear fuels, emphasizing the highest average burn up achieved by these fuel elements. (author)

  12. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    Energy Technology Data Exchange (ETDEWEB)

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly

  13. Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jaluvka, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Wilson, E. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Foyto, L. P. [Univ. of Missouri, Columbia, MO (United States); Kutikkad, K. [Univ. of Missouri, Columbia, MO (United States); McKibben, J. C. [Univ. of Missouri, Columbia, MO (United States); Peters, N. J. [Univ. of Missouri, Columbia, MO (United States)

    2017-02-01

    This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members in the Research and Test Reactor Department at the Argonne National Laboratory (ANL) and the MURR Facility. MURR LEU conversion is part of an overall effort to develop and qualify high-density fuel within the U.S. High Performance Research Reactor Conversion (USHPRR) program conducted by the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization (M3).

  14. Status of neutron beam utilization at the Dalat nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dien, Nguyen Nhi; Hai, Nguyen Canh [Nuclear Research Institute, Dalat (Viet Nam)

    2003-03-01

    The 500-kW Dalat nuclear research reactor was reconstructed from the USA-made 250-kW TRIGA Mark II reactor. After completion of renovation and upgrading, the reactor has been operating at its nominal power since 1984. The reactor is used mainly for radioisotope production, neutron activation analysis, neutron beam researches and reactor physics study. In the framework of the reconstruction and renovation project of the 1982-1984 period, the reactor core, the control and instrumentation system, the primary and secondary cooling systems, as well as other associated systems were newly designed and installed by the former Soviet Union. Some structures of the reactor, such as the reactor aluminum tank, the graphite reflector, the thermal column, horizontal beam tubes and the radiation concrete shielding have been remained from the previous TRIGA reactor. As a typical configuration of the TRIGA reactor, there are four neutron beam ports, including three radial and one tangential. Besides, there is a large thermal column. Until now only two-neutron beam ports and the thermal column have been utilized. Effective utilization of horizontal experimental channels is one of the important research objectives at the Dalat reactor. The research program on effective utilization of these experimental channels was conducted from 1984. For this purpose, investigations on physical characteristics of the reactor, neutron spectra and fluxes at these channels, safety conditions in their exploitation, etc. have been carried out. The neutron beams, however, have been used only since 1988. The filtered thermal neutron beams at the tangential channel have been extracted using a single crystal silicon filter and mainly used for prompt gamma neutron activation analysis (PGNAA), neutron radiography (NR) and transmission experiments (TE). The filtered quasi-monoenergetic keV neutron beams using neutron filters at the piercing channel have been used for nuclear data measurements, study on

  15. Performance of materials in the component cooling water systems of pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B.S.

    1993-06-01

    The component cooling water (CCW) system provides cooling water to several important loads throughout the plant under all operating conditions. An aging assessment CCW systems in pressurized water reactors (PWRs) was conducted as part of Nuclear Plant Aging Research Program (NPAR) instituted by the US Nuclear Regulatory Commission. This paper presents some of the results on the performances of materials in respect of their application in CCW Systems. All the CCW system failures reported to the Nuclear Plant Reliability Data System (NPRDS) from January 1988 to June 1990 were reviewed; it is concluded that three of the main contributors to CCW system failures are valves, pumps, and heat exchangers. This study identified the modes and causes of failure for these components; most of the causes for the aging-related failures could be related to the performance of materials. Also, in this paper the materials used for these components are reviewed, and there aging mechanisms under CCW system conditions are discussed.

  16. Study of a sequencing batch reactor performance in soft drink wastewater treatment

    Directory of Open Access Journals (Sweden)

    Francisco Javier Cuba Terán

    2009-08-01

    Full Text Available A sequencing batch aerobic reactor in pilot scale was constructed and operated with intermittent aeration in Wastewater Treatment Lab of Faculdade de Ciências e Tecnologia de Unesp at Presidente Prudente city. Research was conducted in order to improve reactor’s performance in organic matter and nitrogen removal by means of the application of different aeration times. In 12 and 14 hours long batch tests, with 6 and 8 hours of aeration, more than 96% of organic matter was removed by the third hour in both cases, in the other hand, nitrification showed 50 and 55% of removal at the end of every cycle. Tough showing nitrate removal, denitrification requires more research to be done in order to obtain more accurate data related with best cycle time for both pollutants removal.Key-words: sequencing batchs, aerobic treatment, industrial wastewater.A sequencing batch aerobic reactor in pilot scale was constructed and operated with intermittent aeration in Wastewater Treatment Lab of Faculdade de Ciências e Tecnologia de Unesp at Presidente Prudente city. Research was conducted in order to improve reactor’s performance in organic matter and nitrogen removal by means of the application of different aeration times. In 12 and 14 hours long batch tests, with 6 and 8 hours of aeration, more than 96% of organic matter was removed by the third hour in both cases, in the other hand, nitrification showed 50 and 55% of removal at the end of every cycle. Tough showing nitrate removal, denitrification requires more research to be done in order to obtain more accurate data related with best cycle time for both pollutants removal.Key-words: sequencing batchs, aerobic treatment, industrial wastewater.

  17. Decommissioning of the ASTRA research reactor: Dismantling the auxiliary systems and clearance and reuse of the buildings

    Directory of Open Access Journals (Sweden)

    Meyer Franz

    2008-01-01

    Full Text Available The paper presents work performed in the last phase of the decommissioning of the ASTRA research reactor at the Austrian Research Centers Seibersdorf. Dismantling the pump room installations and the ventilation system, as well as the clearance of the buildings is described. Some conclusions and summary data regarding the timetable, material management, and the cost of the entire project are also presented.

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

    Science.gov (United States)

    Turinsky, Paul J.; Martin, William R.

    2017-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-08-15

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

  20. Reactor in search for money. Cooled neutrons for unique research; Reactor zoekt geld. Gekoelde neutronen maken onderzoek mogelijk

    Energy Technology Data Exchange (ETDEWEB)

    Verdult, E.

    2010-12-17

    The modernization of radiation research at the Delft University of Technology depends on subsidies for new instruments. OYSTER (Optimised Yield for Science, Technology and Education of Radiation) is the plan of the Reactor Institute Delft (RID) to realize such a modernization. The article comprises detailed drawings of the inside of the reactor and illustrates the CNIF (Cold Neutron Irradiation Facility) to fight cancer and POSH-PALS (Positron Annihilation Lifetime Spectrometry) to visualize the atomic structure of materials. [Dutch] De modernisering van het stralingsonderzoek aan de Technische Universiteit Delft staat of valt met subsidie voor een nieuwe opzet van het instrumentarium. OYSTER (Optimised Yield for Science, Technology and Education of Radiation) is het plan van het Reactor Instituut Delft (RID) om de installatie te moderniseren. Het artikel bevat gedetailleerde tekeningen van de binnenkant van de reactor en illustreert de CNIF (Cold Neutron Irradiation Facility) voor de bestrijding van kanker en POSH-PALS (Positron Annihilation Lifetime Spectrometry) waarmee de structuur van materialen op atoomniveau inzichtelijk kan worden gemaakt.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-06-15

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

  3. Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mueller, Don [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-12-01

    In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-06-23

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Beatty, Randy L [ORNL; Harrison, Thomas J [ORNL

    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.

  8. McCARD for Neutronics Design and Analysis of Research Reactor Cores

    Science.gov (United States)

    Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang

    2014-06-01

    McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO, and YALINA.

  9. Feasibility study of fuel cladding performance for application in ultra-long cycle fast reactor

    Science.gov (United States)

    Jung, Ju Ang; Kim, Seung Hyun; Shin, Sang Hun; Bang, In Cheol; Kim, Ji Hyun

    2013-09-01

    As a part of the research and development activities for long-life core sodium-cooled fast reactors, the cladding performance of the ultra-long cycle fast reactor (UCFR) is evaluated with two design power levels (1000 MWe and 100 MWe) and cladding peak temperatures (873 K and 923 K). The key design concept of the UCFR is that it is non-refueling during its 30-60 years of operation. This concept may require a maximum peak cladding temperature of 923 K and a cladding radiation damage of over 200 dpa (displacements per atom). Therefore, for the design of the UCFR, deformation due to thermal creep, irradiation creep, and swelling must be taken into consideration through quantitative evaluations. As candidate cladding materials for use in UCFRs, ferritic-martensitic (FM) steels, oxide dispersion strengthened (ODS) steels, and SiC-based composite materials are studied using deformation behavior modeling for a feasibility evaluation. The results of this study indicate that SiC is a potential UCFR cladding material, with the exception of irradiation creep due to high neutron fluence stemming from its long operating time of about 30-60 years.

  10. Performance Measurement of Research Activitities

    DEFF Research Database (Denmark)

    Jakobsen, Morten; Jensen, Tina Blegind; Peyton, Margit Malmmose

    Performance measurements have made their entry into the world of universities. Every research activity is registered in a database and the output measures form the foundation for managerial decisions. The purpose of this chapter is to investigate the registration practices among researchers...... to the initial registration inscriptions and thus make decisions based on aggregated data that have been cut off from the original phenomena....

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  12. Moving bed biofilm reactor technology: process applications, design, and performance.

    Science.gov (United States)

    McQuarrie, James P; Boltz, Joshua P

    2011-06-01

    The moving bed biofilm reactor (MBBR) can operate as a 2- (anoxic) or 3-(aerobic) phase system with buoyant free-moving plastic biofilm carriers. These systems can be used for municipal and industrial wastewater treatment, aquaculture, potable water denitrification, and, in roughing, secondary, tertiary, and sidestream applications. The system includes a submerged biofilm reactor and liquid-solids separation unit. The MBBR process benefits include the following: (1) capacity to meet treatment objectives similar to activated sludge systems with respect to carbon-oxidation and nitrogen removal, but requires a smaller tank volume than a clarifier-coupled activated sludge system; (2) biomass retention is clarifier-independent and solids loading to the liquid-solids separation unit is reduced significantly when compared with activated sludge systems; (3) the MBBR is a continuous-flow process that does not require a special operational cycle for biofilm thickness, L(F), control (e.g., biologically active filter backwashing); and (4) liquid-solids separation can be achieved with a variety of processes, including conventional and compact high-rate processes. Information related to system design is fragmented and poorly documented. This paper seeks to address this issue by summarizing state-of-the art MBBR design procedures and providing the reader with an overview of some commercially available systems and their components.

  13. The Effect of Duct Level on the Performance of Reactor Vault Cooling System in the PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Sujin; Ryu, Seung Ho; Kim, Dehee; Lee, Tae-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Development of the prototype gen-Ⅵ sodium-cooled fast reactor (PGSFR) has been ongoing in Korea Atomic Energy Research Institute (KAERI). A reactor vault cooling system (RVCS), one of passive decay heat removal systems (PDHRS), passively removes core decay heat by chimney effect when severe accidents occur. The air cooling path is located around containment vessel (CV). An air separator which divides the downstream air and the upstream air is installed between CV and the concrete wall. To design the RVCS, key design parameters such as stack height, gap size between the concrete wall and the air separator, gap size between the air separator and the CV, thickness and layer composition of the air separator have to be determined. A duct level is one of these design parameters. It denotes the height of the upstream air path and related to the heat transfer length from CV to air. The duct level should be optimized with considering structural reliability and heat removal performance. Thus, in this paper, the heat removal performance of RVCS is evaluated depends on the duct level using 1D system design code, that is developed by KAERI autonomously, and commercial CFD program for optimum design of RVCS In this paper, the heat removal performance of RVCS is evaluated depends on the duct level using PARS2- LMR code and commercial CFD program for optimum design of RVCS to satisfy both conflicting needs, structural reliability and cooling performance. As a result of PARS2-LMR code analysis, it was observed that the heat removal rate increases as increase of duct level and the geometrical conditions, that satisfy the design limitations, were obtained. To qualitatively observe the trends of local temperature distribution, CFD simulations were conducted and hotspots were observed at the upper region of ducts for the low duct level case.

  14. Performance of continuous stirred tank reactor (CSTR) on fermentative biohydrogen production from melon waste

    Science.gov (United States)

    Cahyari, K.; Sarto; Syamsiah, S.; Prasetya, A.

    2016-11-01

    This research was meant to investigate performance of continuous stirred tank reactor (CSTR) as bioreactor for producing biohydrogen from melon waste through dark fermentation method. Melon waste are commonly generated from agricultural processing stages i.e. cultivation, post-harvesting, industrial processing, and transportation. It accounted for more than 50% of total harvested fruit. Feedstock of melon waste was fed regularly to CSTR according to organic loading rate at value 1.2 - 3.6 g VS/ (l.d). Optimum condition was achieved at OLR 2.4 g VS/ (l.d) with the highest total gas volume 196 ml STP. Implication of higher OLR value is reduction of total gas volume due to accumulation of acids (pH 4.0), and lower substrate volatile solid removal. In summary, application of this method might valorize melon waste and generates renewable energy sources.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-04-15

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

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

    Directory of Open Access Journals (Sweden)

    Mirmohammadi Sayedeh Leila

    2013-01-01

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

  18. Water chemistry management of research reactor in JAERI

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  19. Irradiation of electronic components and circuits at the Portuguese Research Reactor: Lessons learned

    Energy Technology Data Exchange (ETDEWEB)

    Marques, J.G.; Ramos, A.R.; Fernandes, A.C.; Santos, J.P. [Centro de Ciencias e Tecnologias Nucleares, Instituto Superior Tecnico, Universidade de Lisboa, Estrada Nacional 10, 2695-066 Bobadela LRS (Portugal)

    2015-07-01

    The behavior of electronic components and circuits under radiation is a concern shared by the nuclear industry, the space community and the high-energy physics community. Standard commercial components are used as much as possible instead of radiation hard components, since they are easier to obtain and allow a significant reduction of costs. However, these standard components need to be tested in order to determine their radiation tolerance. The Portuguese Research Reactor (RPI) is a 1 MW pool-type reactor, operating since 1961. The irradiation of electronic components and circuits is one area where a 1 MW reactor can be competitive, since the fast neutron fluences required for testing are in most cases well below 10{sup 16} n/cm{sup 2}. A program was started in 1999 to test electronics components and circuits for the LHC facility at CERN, initially using a dedicated in-pool irradiation device and later a beam line with tailored neutron and gamma filters. Neutron filters are essential to reduce the intensity of the thermal neutron flux, which does not produce significant defects in electronic components but produces unwanted radiation from activation of contacts and packages of integrated circuits and also of the printed circuit boards. In irradiations performed within the line-of-sight of the core of a fission reactor there is simultaneous gamma radiation which complicates testing in some cases. Filters can be used to reduce its importance and separate testing with a pure gamma radiation source can contribute to clarify some irradiation results. Practice has shown the need to introduce several improvements to the procedures and facilities over the years. We will review improvements done in the following areas: - Optimization of neutron and gamma filters; - Dosimetry procedures in mixed neutron / gamma fields; - Determination of hardness parameter and 1 MeV-equivalent neutron fluence; - Temperature measurement and control during irradiation; - Follow-up of reactor

  20. An Expert System to Analyze Homogeneity in Fuel Element Plates for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tolosa, S.C.; Marajofsky, A.

    2004-10-06

    In the manufacturing control of Fuel Element Plates for Research Reactors, one of the problems to be addressed is how to determine the U-density homogeneity in a fuel plate and how to obtain qualitative and quantitative information in order to establish acceptance or rejection criteria for such, as well as carrying out the quality follow-up. This paper is aimed at developing computing software which implements an Unsupervised Competitive Learning Neural Network for the acknowledgment of regions belonging to a digitalized gray scale image. This program is applied to x-ray images. These images are generated when the x-ray beams go through a fuel plate of approximately 60 cm x 8 cm x 0.1 cm thick. A Nuclear Fuel Element for Research Reactors usually consists of 18 to 22 of these plates, positioned in parallel, in an arrangement of 8 x 7 cm. Carrying out the inspection of the digitalized x-ray image, the neural network detects regions with different luminous densities corresponding to U-densities in the fuel plate. This is used in quality control to detect failures and verify acceptance criteria depending on the homogeneity of the plate. This modality of inspection is important as it allows the performance of non-destructive measurements and the automatic generation of the map of U-relative densities of the fuel plate.

  1. Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. A. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Wilson, E. H. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Foyto, L. P. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Kutikkad, K. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; McKibben, J. C. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Peters, N. J. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Cowherd, W. M. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program; Rickman, B. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program

    2014-12-01

    This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo

  2. Measurements and analysis of critical assemblies for research reactors with mixed enrichments

    Energy Technology Data Exchange (ETDEWEB)

    Deen, J.R.; Snelgrove, J.L.; Hobbs, R.W.

    1982-01-01

    As part of the RERTR Program whole-core demonstration in the Ford Nuclear Reactor (FNR) at the University of Michigan, data have been obtained which will allow more extensive validation of neutronics methods for whole-core calculations of an equilibrium high-enriched-uranium (HEU) core and a fresh low-enriched-uranium (LEU) core. A series of experiments designed to provide the data needed for these validations has been performed in the Pool Critical Assembly (PCA) at the Oak Ridge National laboratory (ORNL). This paper reports the results of the measurements and of the subsequent validation calculations performed at ANL. Measurements were made on approximately 20 different critical configurations in the PCA during the period June 15 to 26, 1981. The normal PCA fuel elements contained high-enriched uranium (HEU, 93 wt% /sup 235/U) while the reduced-enrichment fuel elements, obtained for irradiation testing in the Oak Ridge Research Reactor (ORR) under the fuel demonstration activity of the RERTR Program, contained either medium-enriched uranium (MEU, 45 wt% /sup 235/U) or low-enriched uranium (LEU, 19.8 wt% /sup 235/U).

  3. Experience on wet storage spent fuel sipping at IEA-R1 Brazilian research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perrotta, J.A.; Terremoto, L.A.A.; Zeituni, C.A. [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil). Divisao de Engenharia do Nucleo

    1997-12-01

    The IEA-R1 research reactor of the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) is a pool type reactor of B and W design, that has been operating since 1957 at a power of 2 MW. Irradiated (spent) fuels have been stored at the facility during the various years of operation. At present there are 40 spent fuel assemblies at dry storage, 79 spent fuel assemblies at wet storage and 30 fuel assemblies in the core. The oldest fuels are of United States origin, made with U-Al alloy, both of LEU and HEU MTR fuel type. many of these fuel assemblies have corrosion pits along their lateral fuel plates. These pits originate by galvanic corrosion between the fuel plate and the stainless steel storage racks. As a consequence of the possibility of sending the irradiated old fuels back to the U.S.A., sipping tests were performed with the spent fuel assemblies. The reason for this was to evaluate their {sup 137}Cs leaking rate, if any. This work describes the procedure and methodology used to perform the sipping tests with the fuel assemblies at the storage pool, and presents the results obtained for the {sup 137}Cs sipping water activity for each fuel assembly. A correlation is made between the corrosion pits and the activity values measured. A {sup 137}Cs leaking rate is determined and compared to the criteria established for canning spent fuel assemblies before shipment. (author).

  4. Experience on wet storage spent fuel sipping at IEA-R1 Brazilian research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perrotta, J.A.; Terremoto, L.A.A.; Zeituni, C.A

    1998-03-01

    The IEA-R1 research reactor of the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) is a pool type reactor of B and W design, that has been operating since 1957 at a power of 2 MW. Irradiated (spent) fuels have been stored at the facility during the various years of operation. At present there are 40 spent fuel assemblies at dry storage, 79 spent fuel assemblies at wet storage and 30 fuel assemblies in the core. The oldest fuels are of United States origin, made with U-Al alloy, both of LEU and HEU MTR fuel type. Many of these fuel assemblies have corrosion pits along their lateral fuel plates. These pits originate by galvanic corrosion between the fuel plate and the stainless steel storage racks. As a consequence of the possibility of sending the irradiated old fuels back the U.S.A., sipping tests were performed with the spent fuel assemblies. The reason for this was to evaluate their {sup 137}Cs leaking rate, if any. This work describes the procedure and methodology used to perform the sipping tests with the fuel assemblies at the storage pool, and presents the results obtained for the {sup 137}Cs sipping water activity for each fuel assembly. A correlation is made between the corrosion pits and the activity values measured. A {sup 137}Cs leaking rate is determined and compared to the criteria established for canning spent fuel assemblies before shipment.

  5. Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Adorni, Martina [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy)]. E-mail: martina_adorni@tin.it; Bousbia-Salah, Anis [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy); Hamidouche, Tewfik [Commissariat a l' Energie Atomique, Centre de Recherche Nucleaire d' Alger-Algeria, 02 Boulevard Frantz fanon, BP 399 Alger-gare (Algeria); Maro, Beniamino Di [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy); Pierro, Franco [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy); D' Auria, Francesco [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy)

    2005-10-15

    The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool type Research Reactor [IAEA-TECDOC-233, 1980. IAEA Research Reactor Core Conversion from the use of high-enriched uranium to the use of low enriched uranium fuels Guidebook] under flow blockage of a single Fuel Assembly (FA) conditions. Such event was rarely investigated in the open literature notwithstanding the fat that it constitutes a severe accident that may lead to local dryout and eventually to loss of the FA integrity. The transients herein considered are related to partial and total obstruction of the cooling channel of a single Fuel Assembly of the reactor core. This study constitutes the first step of a larger work, which consists in performing a 3D simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic approach of the used thermal-hydraulic system code.

  6. Effect of post-digestion temperature on serial CSTR biogas reactor performance

    DEFF Research Database (Denmark)

    Boe, Kanokwan; Karakashev, Dimitar Borisov; Trably, Eric

    2009-01-01

    The effect of post-digestion temperature on a lab-scale serial continuous-flow stirred tank reactor (CSTR) system performance was investigated. The system consisted of a main reactor operated at 55 degrees C with hydraulic retention time (HRT) of 15 days followed by post-digestion reactors with HRT...... of 5.3 days. Three post-digestion temperatures (55 degrees C, 37 degrees C and 15 degrees C) were compared in terms of biogas production, process stability, microbial community and methanogenic activity, The results showed that the post-digesters operated at 55 degrees C, 37 degrees C and 15 degrees C...... gave extra biogas production of 11.7%, 8.4% and 1.2%, respectively. The post-digester operated at 55 degrees C had the highest biogas production and was the most stable in terms of low VFA concentrations. The specific methanogenic activity tests revealed that the main reactor and the post-digester...

  7. Performance and characterization of a newly developed self-agitated anaerobic reactor with biological desulfurization.

    Science.gov (United States)

    Kobayashi, Takuro; Li, Yu-You

    2011-05-01

    The continuous operation of a newly developed methane fermentation reactor, which requires no electricity for the agitation of the fermentation liquid was investigated, and the extent of the biological desulfurization was monitored. Inside the reactor, the continual change in the liquid level and the self-agitation, occurring between 5 and 16 times every day, distributed the organic load near the inlet port of the reactor, as well as providing a nutrient supply to the hydrogen sulfide oxidizing bacteria. At different COD(Cr) loading rates (5, 7, 10 kg m(3)d(-1)), the reactor achieved a biogas production yield of 0.72-0.82 m(3)g(-1)-TS, a COD(Cr) reduction of 79.4-85.5% and an average of 99% hydrogen sulfide removal. This investigation demonstrated that the self-agitated reactor is comparable in digestion performance to the completely stirred tank reactor (CSTR) investigated in a previous study, and that the desulfurization performance was significantly enhanced compared to the CSTR.

  8. Study on the flow characteristics and the wastewater treatment performance in modified internal circulation reactor.

    Science.gov (United States)

    Wang, Jiade; Xu, Weijun; Yan, Jingjia; Yu, Jianming

    2014-12-01

    A modified internal circulation (MIC) reactor with an external circulation system was proposed and the performance of treating dyeing wastewater using both MIC and typical IC reactor were compared. Utilization of the external circulation system in the MIC reactor could dramatically improve the mixing intensity of the biomass with the wastewater and resulted in better performance. The COD removal efficiency, biogas production, volatile fatty acids and effluent color were approximately 87%, 98 L d−1, 180 mg L−1 and 100 times, respectively, in the MIC reactor with a hydraulic retention time of 5 h and organic loading rate of 15 kg COD m−3 d−1. The hydrodynamics of the MIC reactor under different flows rate of external circulation were also analyzed using computational fluid dynamics (CFD) method. The optimal flow rate of external circulation was 12 L min−1, which resulted in a corresponding up-flow velocity of 40 m h−1. The consistency of the result between experiment and simulation validated the scientificity of CFD technique applied to numerical simulation of the MIC reactor.

  9. DISMANTLING OF THE UPPER RPV COMPONENTS OF THE KARLSRUHE MULTI-PURPOSE RESEARCH REACTOR (MZFR), GERMANY

    Energy Technology Data Exchange (ETDEWEB)

    Prechtl, E.; Suessdorf, W.

    2003-02-27

    The Multi-purpose Research Reactor was a pressurized-water reactor cooled and moderated with heavy water. It was built from 1961 to 1966 and went critical for the first time on 29 September 1965. After nineteen years of successful operation, the reactor was de-activated on 3 May 1984. The reactor had a thermal output of 200 MW and an electrical output of 50 MW. The MZFR not only served to supply electrical power, but also as a test bed for: - research into various materials for reactor building (e. g. zirkaloy), - the manufacturing and operating industry to gain experience in erection and operation, - training scientific and technical reactor staff, and - power supply (first nuclear combined-heat-and-power system, 1979-1984). The experience gained in operating the MZFR was very helpful for the development and operation of power reactors. At first, safe containment and enclosure of the plant was planned, but then it was decided to dismantle the plant completely, step by step, in view o f the clear advantages of this approach. The decommissioning concept for the complete elimination of the plant down to a green-field site provides for eight steps. A separate decommissioning license is required for each step. As part of the dismantling, about 72,000 Mg [metric tons] of concrete and 7,200 Mg of metal (400 Mg RPV) must be removed. About 700 Mg of concrete (500 Mg biological shield) and 1300 Mg of metal must be classified as radioactive waste.

  10. Gender inequality and research performance

    DEFF Research Database (Denmark)

    Nielsen, Mathias Wullum

    2016-01-01

    inherent to the academic system or instead individualistic matters, such as personal motivation, performance and merit. In order to contribute to a more nuanced understanding of this particular issue, this cross-sectional bibliometric study investigates the link between gender and research performance......Academic debates addressing the persistent gender gap in science reveal considerable contestation of the relevance and extent of the problem. Particular attention has been given to the question of whether women’s high attrition rates should be ascribed to the structural and cultural barriers...... performance gap in favour of male researchers. The result has implications for research organizations and managers, as it raises concerns about the validity of individual-meritocratic explanations of the skewed gender distributions in academia....

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

    OpenAIRE

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

    2013-01-01

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

  12. Control rod reactivity worth determination of a typical MTR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rizwan, M.; Raza, S.S.; Khan, R. [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad (Pakistan). Dept. of Nuclear Engineering

    2015-10-15

    The safe and reliable utilization of research reactor demands the possible accurate information of control rod (CR) worths. The criticality positions of the control rods changes with time due to build up fission products. It is therefore important to determine the reactivity worth of control rods. The aim of this article is to estimate the reactivity worth of controls rods in the equilibrium core of a Materials Testing Reactor (MTR). A deterministic model of the reactor core was developed and confirmed against the reference results of excess reactivity, shutdown margin and combined control rod reactivity worth using the combination of WIMS/D4 and CITATION computer codes.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Klupak, V.; Viererbl, L.; Lahodova, Z.; Marek, M.; Vins, M. [Research Centre Rez Ltd., Husinec-Rez 130 (Czech Republic)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

  18. Research and development program for PWR safety at the CEA reactor thermal hydraulics laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, M. [CEA, Grenoble (France)

    1995-04-15

    Since the start of the French electronuclear program, the three partners Fermate, EDF and Cea (DRN and IPSN) have devoted considerable effort to research and development for safety issues. In particular an important program on thermal hydraulics was initiated at the beginning of the seventies. It is illustrated by the development of the CATHARE thermalhydraulic safety code which includes physical models derived from a large experimental support program and the construction of the BETHSY integral facility which is aimed to assess both the CATHARE code and the physical relevance of the accident management procedures to be applied on reactors. The state of the art on this program is described with particular emphasis on the capabilities and the assessment of the last version of CATHARE and the lessons drawn from 50 BETHSY tests performed so far. The future plans for safety research cover the following strategy: - to solve the few problems identified on present computing tools and extend the assessment - to solve the few problems identified on present computing tools and extend the assessment - to perform safety studies on the basis of plant operation feedback - to contribute to treating the safety issues related to the future reactors and in particular the case of severe accidents which have to be taken into account from the design stage. The program on severe accidents is aimed to support the design studies performed by the industrial partners and to provide computing tools which model the various phases of severe accidents and will be validated on experiments performed with real and simulating materials. The main lines of the program are: - the development of the TOLBIAC 3D code for the thermal hydraulics of core melt pools, which will be validated against the Bali experiment presently under construction - the Sultan experiment, to study the capability of cooling by external flooding of the reactor vessel - the development of the MC-3D code for core melt

  19. Evaluate the radioactivity along the central thimble hole of a decommissioned heavy water research reactor using TLD approach.

    Science.gov (United States)

    Lee, Lun-Hui; Sher, Hai-Feng; Lu, I-Hsin; Pan, Lung-Kwang

    2012-04-01

    The radioactivity along the central thimble hole of a decommissioned heavy water research reactor, TRR, was evaluated using TLD approach. The decay radionuclide was verified to be Co-60. The dose along the TRR central thimble hole was detected and revised by performing an unfolding analysis. The revised data reduced to 70-90% of the original data (for example, the maximum dose rate was reduced from 6447 to 4831 mSv/h,) and were more reliable.

  20. Hydraulic performance of a proposed in situ photocatalytic reactor for degradation of MTBE in water.

    Science.gov (United States)

    Lim, Leonard Lik Pueh; Lynch, Rod

    2011-01-01

    Methyl tert-butyl ether (MTBE) groundwater remediation projects often require a combination of technologies resulting in increasing the project costs. A cost-effective in situ photocatalytic reactor design, Honeycomb II, is proposed and tested for its efficiency in MTBE degradation at various flows. This study is an intermediate phase of the research in developing an in situ photocatalytic reactor for groundwater remediation. It examines the effect of the operating variables: air and water flow and double passages through Honeycomb II, on the MTBE removal. MTBE vaporisation is affected by not only temperature, Henry's law constant and air flow to volume ratio but also reactor geometry. The column reactor achieved more than 84% MTBE removal after 8 h at flows equivalent to horizontal groundwater velocities slower than 21.2 cm d⁻¹. Despite the contrasting properties between a photocatalytic indicator methylene blue and MTBE, the reactor efficiency in degrading both compounds showed similar responses towards flow (equivalent groundwater velocity and hydraulic residence time (HRT)). The critical HRT for both compounds was approximately 1 d, which corresponded to a velocity of 21.2 cm d⁻¹. A double pass through both new and used catalysts achieved more than 95% MTBE removal after two passes in 48 h. It also verified that the removal efficiency can be estimated via the sequential order of the removal efficiency of one pass obtained in the laboratory. This study reinforces the potential of this reactor design for in situ groundwater remediation.

  1. Radiation protection tasks on the Kiev research reactor WWR-M

    Directory of Open Access Journals (Sweden)

    Lobach Yuri N.

    2009-01-01

    Full Text Available Both the description of and the operational experience with the radiation protection system at the research reactor WWR-M are presented. The list of the factors regarding the radiation hazards during the reactor routine operation is given and the main activities on the radiation safety provision are established. The statistical information for the staff exposure, the radioactive aerosol releases and the external radiation monitoring is shown. The preliminary considerations on the system upgrading for the decommissioning are presented.

  2. Influence of alkalinity and VFAs on the performance of an UASB reactor with recirculation for the treatment of Tequila vinasses.

    Science.gov (United States)

    López-López, Alberto; León-Becerril, Elizabeth; Rosales-Contreras, María Elena; Villegas-García, Edgardo

    2015-01-01

    The main problem linked to the stability of upflow anaerobic sludge blanket (UASB) reactors during the treatment of Tequila vinasse is the high acidity and the null alkalinity present in this effluent. This research evaluates the effect of alkalinity and volatile fatty acids (VFAs) concentration on the performance of an UASB reactor with recirculation of the effluent for removing organic matter and biogas production from Tequila vinasses. Recirculation of the effluent reduces the impact of VFAs and organic matter concentration present in the influent, inducing the stability of the reactor. The UASB reactor was operated during 235 days at organic loading rates from 2.5 to 20.0 kg m(-3) d(-1), attaining a removal efficiency of COD greater than 75% with a methane yield of 335 ml CH4 g(-1) COD at SPT, maintaining a ratio of VFAs/Alk ≤ 0.5. Therefore, an optimal ratio of VFAs/Alk was established for the system operating in stable conditions for the treatment of Tequila vinasses. Under these conditions, the alkalinity was recuperated by the system itself, without the addition of external alkalinity.

  3. Experimental study of flow inversion in MTR upward flow research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Abdel-Hadi, Ead A. [Benha Univ., Cairo (Egypt). Shobra Faculty of Engineering; Khedr, Ahmed; Talha, Kamal Eldin Aly; Abdel-Latif, Salwa Helmy

    2014-06-15

    The core cooling of upward flow MTR pool type Research Reactor (RR) at the later stage of pump coast down is experimentally handled to clarify the effect of some operating parameters on RR core cooling. Therefore, a test rig is designed and built to simulate the core cooling loop at this stage. The core is simulated as two vertical channels, electrically heated, and extended between upper and lower plenums. Two elevated tanks filled with water are connected to the two plenums. The first one constitutes a left branch, connected to the lower plenum, and is electrically heated to simulate the core return pipe. The second one constitutes the right branch, connected to the upper plenum, and is cooled by refrigerant circuit to simulate the reactor pool. Channel coolant and wall temperatures at different power and branch temperatures are measured, registered and analyzed. The results show that at this stage of core cooling two cooling loops are established; an internal circulation loop between the channels dominated by the difference in channel's power and an external circulation loop between the branches dominated by the temperature difference between branches. Also, there is a double inversion in core flow, upward-downward-upward flow. This double inversion increases largely the channel's wall temperature. Complementary safety analysis to evaluate this phenomenon must be performed. (orig.)

  4. A neutronic feasibility study for LEU conversion of the WWR-SM research reactor in Uzbekistan.

    Energy Technology Data Exchange (ETDEWEB)

    Rakhmanov, A.

    1998-10-19

    The WWR-SM research reactor in Uzbekistan has operated at 10 MW since 1979, using Russian-supplied IRT-3M fuel assemblies containing 90% enriched uranium. Burnup tests of three full-sized IRT-3M FA with 36% enrichment were successfully completed to a burn up of about {approximately}50% in 1987-1989. In August 1998, four IRT-3M FA with 36% enriched uranium were loaded into the core to initiate conversion of the entire core to 36% enriched fuel. This paper presents the results of equilibrium fuel cycle comparisons of the reactor using HEU (90%) and HEU (36%) IRT-3M fuel and compares results with the performance of IRT-4M FA containing LEU (19.75%). The results show that an LEU (19.75%) density of 3.8 g/cm{sup 3} is required to match the cycle length of the HEU (90%) core and an LEU density 3.9 g/cm{sup 3} is needed to match the cycle length of the HEU (36%) core.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-02-01

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

  6. Study on the effect of the impeller and diffuser blade number on reactor coolant pump performances

    Science.gov (United States)

    Long, Y.; Yin, J. L.; Wang, D. Z.; Li, T. B.

    2016-05-01

    In this paper, CFD approach was employed to study how the blade number of impeller and diffuser influences reactor coolant pump performances. The three-dimensional pump internal flow channel was modelled by pro/E software, Reynolds-averaged Naiver-Stokes equations with the k-ε turbulence model were solved by the computational fluid dynamics software CFX. By post-processing on the numerical results, the performance curves of reactor coolant pump were obtained. The results are as follows, with the blade number of the impeller increasing, the head of the pump with different diffuser universally increases in the 8Q n∼1.2Q n conditions, and at different blade number of the diffuser, the head increases with the blade number of the impeller increasing. In 1.0Q n condition, when the blades number combination of impeller and diffuser chooses 4+16, 7+14 and 6+18, the head curves exist singular points. In 1.2Q n condition, the head curve still exists singular point in 6+18. With the blade number of the impeller increasing, the efficiency of the pump with different diffuser universally decreases in the 0.8Q n and 1.0Q n conditions, but in 1.2Q n condition, the efficiency of the pump with different diffuser universally increases. In 1.0Q n condition, the impellers of 4 and 5 blades are better. When the blade number combination of impeller and diffuser choose 4+11, 4+17, 4+18, 5+12, 5+17 and 5+18, the efficiencies relatively have higher values. With the blade number of the impeller increasing, the hydraulic shaft power of the pump with different diffuser universally increases in the 0.8Q n∼1.2Q n conditions, and with the blade number of the diffuser increasing, the power of different impeller overall has small fluctuation, but tends to be uniform. This means the increase of the diffuser blade number has less influence on shaft power.The influence on the head and flow by the matching relationship of the blades number between impeller and diffuser is very complicated, which

  7. Disposal of irradiated fuel elements from German research reactors. Status and outlook

    Energy Technology Data Exchange (ETDEWEB)

    Thamm, G. [Central Research Reactor and Nuclear Operations Division, Research Centre Juelich, Forschungszentrum Juelich GmbH, Juelich (Germany)

    1999-07-01

    There will be a quantity of highly radioactive spent nuclear fuel (snf) from German research reactors amounting to about 9.1 t by the end of the next decade, which has to be disposed of. About 4.1 t of this quantity are intended to be returned to the USA. The remaining approximately 5 t can be loaded into approximately 30 CASTOR-2 casks and will be stored in a central German dry interim store for about 30 to 50 years (first step of the domestic disposal concept). Of course, snf arising from the operation of research reactors beyond 2010 has to be disposed of in the same way (3 MTR-2 casks every two years for BER-II and FRM-II). It is expected that snf from the zero-power facilities probably will be recycled for reusing the uranium. Due to the amendment of the German Atomic Energy Act intended by the new Federal German Government, the interim dry storage of snf from power reactors in central storage facilities like Ahaus or Gorleben will be stopped and the power reactors have to store snf at their own sites. Although the amendment only concerns nuclear power reactors, it could not be excluded that snf from research reactors, too, cannot be stored at Ahaus or Gorleben at present. (author)

  8. Impact of asymmetric lamp positioning on the performance of a closed-conduit UV reactor

    Directory of Open Access Journals (Sweden)

    Tipu Sultan

    2017-06-01

    Full Text Available Computational fluid dynamics (CFD analyses for the performance improvement of a closed-conduit ultraviolet (UV reactor were performed by changing the lamp positions from symmetric to asymmetric. The asymmetric lamp positioning can be useful for UV reactor design and optimization. This goal was achieved by incorporating the two performance factors, namely reduction equivalent dose (RED and system dose performance. Four cases were carried out for asymmetric lamp positioning within the UV reactor chamber and each case consisted of four UV lamps that were simulated once symmetrically and four times asymmetrically. The results of the four asymmetric cases were compared with the symmetric one. Moreover, these results were evaluated by using CFD simulations of a closed-conduit UV reactor. The fluence rate model, UVCalc3D was employed to validate the simulations results. The simulation results provide detailed information about the dose distribution, pathogen track modeling and RED. The RED value was increased by approximately 15% by using UVCalc3D fluence rate model. Additionally, the asymmetric lamp positioning of the UV lamps had more than 50% of the pathogens received a better and a higher UV dose than in the symmetric case. Consequently, the system dose performance was improved by asymmetric lamp positioning. It was concluded that the performance parameters (higher RED and system dose performance were improved by using asymmetric lamp positioning.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  10. Research on physical and chemical parameters of coolant in Light-Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

  11. Design considerations for post accident monitoring system of a research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Gwi Sook; Park, Je Yun; Kim, Young Ki [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    The Post Accident Monitoring System (PAMS) provides primary information for operators to assess the plant conditions and perform their role in bringing the plant to a safe condition during an accident. The PAMS of NPP (Nuclear Power Plant) in KOREA provides the continuous display of the PAM category 1 parameters specified in R.G 1.97, Rev. 03. Recently the PAMS of NPP has been designed according to R.G 1.97, Rev. 04. There is no PAMS at the HANARO in KOREA, but recently RRs (Research Reactors) around the world are going to have PAMS for various multi purposes. We should determine the design considerations for PAMS in a Korean RR based on the design state analysis. Thus, this paper proposes strategies on the design considerations for the PAMS of a Korean RR.

  12. Fabrication of DUPIC Fuel for the 6th Irradiation Test at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Ki; Lee, D. Y.; Cho, K. H.; Kim, S. S.; Lee, J. W.; Lee, Jae W.; Park, G. I.; Lee, C. Y.; Yang, M. S

    2006-02-15

    In this study, 15 DUPIC pellets and two mini-elements were fabricated to precisely investigate the thermal characteristics of DUPIC fuel. As a result of the experiment, DUPIC pellets with 10.221{approx}10.278 g/cm{sup 3} (94.5{approx}95.0 % of T.D.) of sintered density and less than Ra 0.96 {mu}m of surface roughness satisfying the specifications of DUPIC fuel for the 6th irradiation test have been remotely fabricated at hot cell. 5 DUPIC pellets including 3 pellets equipped with thermal sensor in the center of the pellet were loaded in a mini-element. Endcap welding of the mini-element was performed by Nd:YAG laser. The DUPIC mini-elements assembled in an instrumented rig will be irradiated at HANARO research reactor.

  13. TRIGA-TRAP: A penning trap mass spectrometer at the research reactor TRIGA Mainz

    Energy Technology Data Exchange (ETDEWEB)

    Smorra, Christian [Physikalisches Institut, Universitaet Heidelberg (Germany); Institut fuer Kernchemie, Universitaet Mainz (Germany); Blaum, Klaus [Physikalisches Institut, Universitaet Heidelberg (Germany); Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany); Block, Michael; Herfurth, Frank [GSI, Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); Eberhardt, Klaus [Institut fuer Kernchemie, Universitaet Mainz (Germany); Eibach, Martin; Ketelaer, Jens; Ketter, Jochen; Knuth, Konstantin; Repp, Julia [Institut fuer Physik, Universitaet Mainz (Germany); Nagy, Szilard [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany)

    2009-07-01

    Nuclear masses represent the binding energies and, therefore, the sum of all interactions in the nucleus. They provide an important input parameter to nuclear structure models. Presently, a tremendous interest in masses of very exotic neutron-rich nuclides exists to support theoretical models for the nucleosynthesis via the rapid neutron capture process. The research reactor TRIGA Mainz provides access to a large variety of neutron-rich nuclides produced by thermal-neutron induced fission of an actinide target. The double-Penning trap mass spectrometer TRIGA-TRAP will perform high-precision mass measurements in this region of the nuclear chart as well as on actinides from uranium to californium. It also serves as a test facility for the development of new techniques that will be implemented in future facilities like MATS at FAIR (GSI, Darmstadt). The layout of TRIGA-TRAP as well as recent mass measurements are presented.

  14. LOFA and RIA analysis of the Indonesian Multipurpose research reactor RSG-GAS (1)

    Energy Technology Data Exchange (ETDEWEB)

    Endiah Puji Hastuti; Hudi Hastowo; Iman Kuntoro [Center for Multipurpose Reactor, National Atomic Energy Agency (BATAN), Puspiptek, Serpong, Tangerang (Indonesia)

    1999-07-01

    Investigation on accident of the Indonesian Multipurpose research reactor RSG-GAS has been performed by computer simulation technique. Two groups of transients were considered, namely transient due to loss of primary cooling system (LOFA) and power excursion due to reactivity insertion (RIA). In such a transient condition, the Common Mode Failure (CMF) is considered and it will induce a situation so called unprotected transient or Anticipated Transient Without Scram (ATWS). RELAP5, PARET-ANL and EUREKA-2RR computer packages have been applied for these analyses. Simulations result done using these computer packages showed that in the occurrence of LOFA and RIA, failure on fuel elements is limited to the region with the highest power factor. (author)

  15. Inspection of domestic nuclear fuel rods using neutron radiography at the Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dastjerdi, Mohammad Hosein Choopan; Khalafi, Hossein; Kasesaz, Yaser [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Movafeghi, Amir

    2016-11-01

    Three unused domestic fuel rods were investigated qualitatively and quantitatively by means of thermal neutron radiography. The neutron radiography tests were performed by the image plate method at Tehran research reactor in order to check the fuel properties. The pellets of these three fuel rods contained three different U-235 enrichments and different sizes that were filled into a zircalloy tube. In the qualitative investigations, the difference in size and enrichment between the pellets and the gaps between them were obviously recognized in the image of the fuel rods. In the quantitative investigations, data of the pellets compositions, their sizes (lengths and diameters) and the gaps between them were extracted from obtained images. It was found that the measured data and the manufacturer's specifications are in good agreement.

  16. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  17. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration (Figure 1). The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration (Figure 2), based on current reactor use, has been defined for the fuel conversion analyses [1]. The code RELAP5/Mod 3.3 [2] was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  18. A comparative assessment of independent thermal-hydraulic models for research reactors: The RSG-GAS case

    Energy Technology Data Exchange (ETDEWEB)

    Chatzidakis, S., E-mail: schatzid@purdue.edu [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907 (United States); Hainoun, A. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400, San Carlos de Bariloche, Rio Negro (Argentina); Alhabet, F. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Francioni, F. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400, San Carlos de Bariloche, Rio Negro (Argentina); Ikonomopoulos, A. [Institute of Nuclear and Radiological Sciences, Energy, Technology and Safety, National Center for Scientific Research ‘Demokritos’, 15130, Aghia Paraskevi, Athens (Greece); Ridikas, D. [Department of Nuclear Sciences and Applications, International Atomic Energy Agency, Vienna International Centre, A-1400 Vienna (Austria)

    2014-03-15

    Highlights: • Increased use of thermal-hydraulic codes requires assessment of important phenomena in RRs. • Three independent modeling teams performed analysis of loss of flow transient. • Purpose of this work is to examine the thermal-hydraulic codes response. • To perform benchmark analysis comparing the different codes with experimental measurements. • To identify the impact of the user effect on the computed results, performed with the same codes. - Abstract: This study presents the comparative assessment of three thermal-hydraulic codes employed to model the Indonesian research reactor (RSG-GAS) and simulate the reactor behavior under steady state and loss of flow transient (LOFT). The RELAP5/MOD3, MERSAT and PARET-ANL thermal-hydraulic codes are used by independent research groups to perform benchmark analysis against measurements of coolant and clad temperatures, conducted on an instrumented fuel element inside RSG-GAS core. The results obtained confirm the applicability of RELAP5/MOD3, MERSAT and PARET-ANL on the modeling of loss of flow transient in research reactors. In particular, the three codes are able to simulate flow reversal from downward forced to upward natural convection after pump trip and successful reactor scram. The benchmark results show that the codes predict maximum clad temperature of hot channel conservatively with a maximum overestimation of 27% for RELAP5/MOD3, 17% for MERSAT and 8% for PARET-ANL. As an additional effort, the impact of user effect on the simulation results has been assessed for the code RELAP5/MOD3, where the main differences among the models are presented and discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

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

  1. Design improvement and performance evaluation of solar photocatalytic reactor for industrial effluent treatment.

    Science.gov (United States)

    Nair, Ranjith G; Bharadwaj, P J; Samdarshi, S K

    2016-12-01

    This work reports the details of the design components and materials used in a linear compound parabolic trough reactor constructed with an aim to use the photocatalyst for solar photocatalytic applications. A compound parabolic trough reactor has been designed and engineered to exploit both UV and visible part of the solar irradiation. The developed compound parabolic trough reactor could receive almost 88% of UV radiation along with a major part of visible radiation. The performance of the reactor has been evaluated in terms of degradation of a probe pollutant using the parameters such as rate constant, residence time and photonic efficiency. An attempt has been made to assess the performance in different ranges of solar spectrum. Finally the developed reactor has been employed for the photocatalytic treatment of a paper mill effluent using Degussa P25 as the photocatalyst. The paper mill effluent collected from Nagaon paper mill, Assam, India has been treated under both batch mode and continuous mode using Degussa P25 photocatalyst under artificial and natural solar radiation, respectively. The photocatalytic degradation kinetics of the paper mill effluent has been determined using the reduction in total organic carbon (TOC) values of the effluent.

  2. Monte Carlo analysis of the accelerator-driven system at Kyoto University Research Reactor Institute

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won Kyeong; Lee, Deok Jung [Nuclear Engineering Division, Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Lee, Hyun Chul [VHTR Technology Development Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Pyeon, Cheol Ho [Nuclear Engineering Science Division, Kyoto University Research Reactor Institute, Osaka (Japan); Shin, Ho Cheol [Core and Fuel Analysis Group, Korea Hydro and Nuclear Power Central Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    Directory of Open Access Journals (Sweden)

    Wonkyeong Kim

    2016-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-07-01

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

  6. End of the line for Harwell's Dido and Pluto research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Tom; Nicholson, K.

    1990-04-01

    After 34 years of continuous operation the Dido and Pluto research reactors were shutdown for the last time on the 31 March 1990. The history of their development and contributions to the UK nuclear programme, isotope production, support to industry and basic scientific research are described. (author).

  7. Decommissioning of the Astra research reactor: Review and status on July 2003

    Directory of Open Access Journals (Sweden)

    Meyer Franz

    2003-01-01

    Full Text Available The paper describes work on the decommissioning of the ASTRA research reactor at the Austrian Research Centers Seibersdorf. Organizational, planning, and dismantling work done until July 2003 including radiation protection and waste management procedures as well as the current status of the project are presented. Completion of the decommissioning activities is planned for 2006.

  8. Advanced reactor safety research quarterly report, October-December 1982. Volume 24

    Energy Technology Data Exchange (ETDEWEB)

    None

    1984-04-01

    This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  10. Performance of commercial off-the-shelf microelectromechanical systems sensors in a pulsed reactor environment

    Energy Technology Data Exchange (ETDEWEB)

    Hobert, Keith Edwin [Los Alamos National Laboratory; Heger, Arlen S [Los Alamos National Laboratory; Mccready, Steven S [Los Alamos National Laboratory

    2010-07-15

    Prompted by the unexpected failure of piezoresistive sensors in both an elevated gamma-ray environment and reactor core pulse tests, we initiated radiation testing of several MEMS piezoresistive accelerometers and pressure transducers to ascertain their radiation hardness. Some commercial off-the-shelf sensors are found to be viable options for use in a high-energy pulsed reactor, but others suffer severe degradation and even catastrophic failure. Although researchers are promoting the use of MEMS devices in radiation-harsh environment, we nevertheless find assurance testing necessary.

  11. Electro-catalytic membrane reactors and the development of bipolar membrane technology

    NARCIS (Netherlands)

    Balster, J.; Stamatialis, D.F.; Wessling, M.

    2004-01-01

    Membrane reactors are currently under extensive research and development. Hardly any concept, however, is realized yet in practice. Frequently, forgotten as membrane reactors are electro-catalytic membrane reactors where electrodes perform chemical conversations and membranes separate the locations

  12. Kinetic evaluation and process performance of an upflow anaerobic filter reactor degrading terephthalic acid.

    Science.gov (United States)

    Davutluoglu, Orkun I; Seckin, Galip

    2014-01-01

    The anaerobic degradation of terephthalic acid (TA) as the sole organic carbon source was studied in an upflow anaerobic filter (UAF) reactor. The reactor was seeded with biomass obtained from a full-scale upflow anaerobic sludge bed (UASB) reactor and was used to treat wastewater from a petrochemical facility producing dimethyl terephthalate. The UAF reactor was operated for 252 d with a constant hydraulic retention time of 24 h, and the organic loading rate (OLR) was gradually increased from 1 to 10 g-chemical oxygen demand (COD)/L d. After a lag period of approximately 40 d, the COD removal efficiency increased exponentially and high removal rate values (≈90%) were obtained, except for at highest OLR (10 g-COD/L d). The high removal rates and the robustness of the reactor performance could be attributed to the formation of biofilm as well as granular sludge. The methane production rates (0.22 to 2.15 L/d) correlated well with the removed OLRs (0.3 to 6.8 g-COD/L d) during the various phases of treatment, indicating that the main mechanism of TA degradation occurs via methanogenic reactions. The average methane content of the produced biogas was 70.3%. The modified Stover-Kincannon model was found to be applicable for the anaerobic degradation of TA in UAFs (Umax = 64.5, KB = 69.1 g-COD/L d and Ymax = 0.27 L-CH4/g-CODremoved). These results suggest that UAF reactors are among the most effective reactor configurations for the anaerobic degradation of TA.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

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

  18. Analysis on approach of safeguards implementation at research reactor handling item count and bulk material

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun Jo; Lee, Sung Ho; Lee, Byung Doo; Jung, Juang [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    KiJang research reactor (KJRR) will be constructed to produce the radioisotope such as Mo-99 etc., provide the neutron transmutation doping (NTD) service of silicon, and develop the core technologies of research reactor. In this paper, the features of the process and nuclear material flow are reviewed and the material balance area (MBA) and key measurement point (KMP) are established based on the nuclear material flow. Also, this paper reviews the approach on safeguards and nuclear material accountancy at the facility level for Safeguards-by-Design at research reactor handling item count and bulk material. In this paper, MBA and KMPs are established through the analysis on facility features and major process at KJRR handling item count and bulk material. Also, this paper reviews the IAEA safeguards implementation and nuclear material accountancy at KJRR. It is necessary to discuss the safeguards approach on the fresh FM target assemblies and remaining uranium in the intermediate level liquid wastes.

  19. Fuel irradiation research of Japan at OECD Halden Reactor Project. Achievement of joint researches between JAERI and other organizations in the period from 1994 to 1996

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi; Nakamura, Jinichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kinoshita, Motoyasu [and others

    1998-01-01

    JAERI has performed cooperative researches with many Japanese agencies and companies by means of the Halden Boiling Heavy Water Reactor (HBWR) which is located at Halden in Norway. These cooperative researches are carried out based on the contracts of the cooperative researches, which are revised every three years, in accordance with the renewal of the participation of JAERI to the OECD Halden Reactor Project. This report summaries the objectives, contents and the outlines of the achievements of the cooperative researches during the three years from 1994 January to 1996 December. During the period, ten cooperative researches had been carried out, and two of them had finished during the period and other eight researches has been continued to the next three year period. There are many research items, and most of them are irradiation test researches of advanced fuel and cladding concerned with the high burnup utilization of LWR fuel or MOX fuel irradiation researches to prepare for the introduction of Plutonium utilization in LWRs. The researches of fuel irradiation usually take long time because of the characteristics of these kind of research work, and three years are usually not enough to obtain some achievements from the irradiation tests. Therefore, eight tests have been continued after the three year period. In this report, the achievements of the continued researches to the next three year period are not final one but a kind of progress report. (author)

  20. Initial in-reactor performance of the Cornell cold neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Spern, S.A.; Clark, D.D.; Atwood, A.G. [Cornell Univ., Ithaca, NY (United States)

    1996-12-31

    The Cornell Cold Neutron Beam Facility consists of two major subsystems, a cold neutron source (CNS) and a 13-m-long curved neutron guide. This paper describes the initial in-reactor performance tests of the CNS. The results agree closely with predictions from out-of-reactor tests and meet the design criteria for safety and simplicity of operation. This phase of the project has therefore been completed. Three meters of neutron guide were in place during these tests, and a preliminary evaluation of neutronic properties is also presented.

  1. A Comparison of Photocatalytic Oxidation Reactor Performance for Spacecraft Cabin Trace Contaminant Control Applications

    Science.gov (United States)

    Perry, Jay L.; Frederick, Kenneth R.; Scott, Joseph P.; Reinermann, Dana N.

    2011-01-01

    Photocatalytic oxidation (PCO) is a maturing process technology that shows potential for spacecraft life support system application. Incorporating PCO into a spacecraft cabin atmosphere revitalization system requires an understanding of basic performance, particularly with regard to partial oxidation product production. Four PCO reactor design concepts have been evaluated for their effectiveness for mineralizing key trace volatile organic com-pounds (VOC) typically observed in crewed spacecraft cabin atmospheres. Mineralization efficiency and selectivity for partial oxidation products are compared for the reactor design concepts. The role of PCO in a spacecraft s life support system architecture is discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Robert Stephen [Idaho National Laboratory

    2015-12-01

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

  3. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary

    Energy Technology Data Exchange (ETDEWEB)

    Shannon Bragg-Sitton

    2014-02-01

    Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

  4. Reviewing real-time performance of nuclear reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Preckshot, G.G. [Lawrence Livermore National Lab., CA (United States)

    1993-08-01

    The purpose of this paper is to recommend regulatory guidance for reviewers examining real-time performance of computer-based safety systems used in nuclear power plants. Three areas of guidance are covered in this report. The first area covers how to determine if, when, and what prototypes should be required of developers to make a convincing demonstration that specific problems have been solved or that performance goals have been met. The second area has recommendations for timing analyses that will prove that the real-time system will meet its safety-imposed deadlines. The third area has description of means for assessing expected or actual real-time performance before, during, and after development is completed. To ensure that the delivered real-time software product meets performance goals, the paper recommends certain types of code-execution and communications scheduling. Technical background is provided in the appendix on methods of timing analysis, scheduling real-time computations, prototyping, real-time software development approaches, modeling and measurement, and real-time operating systems.

  5. Development of an Integrated Performance Model for TRISO-Coated Gas Reactor Particle Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew; Miller, Gregory Kent; Martin, David George; Maki, John Thomas

    2005-05-01

    The success of gas reactors depends upon the safety and quality of the coated particle fuel. The understanding and evaluation of this fuel requires development of an integrated mechanistic fuel performance model that fully describes the mechanical and physico-chemical behavior of the fuel particle under irradiation. Such a model, called PARFUME (PARticle Fuel ModEl), is being developed at the Idaho National Engineering and Environmental Laboratory. PARFUME is based on multi-dimensional finite element modeling of TRISO-coated gas reactor fuel. The goal is to represent all potential failure mechanisms and to incorporate the statistical nature of the fuel. The model is currently focused on carbide, oxide nd oxycarbide uranium fuel kernels, while the coating layers are the classical IPyC/SiC/OPyC. This paper reviews the current status of the mechanical aspects of the model and presents results of calculations for irradiations from the New Production Modular High Temperature Gas Reactor program.

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

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

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

  7. Optimization process for the design of the DCLL blanket for the European DEMOnstration fusion reactor according to its nuclear performances

    Science.gov (United States)

    Palermo, Iole; Rapisarda, David; Fernández-Berceruelo, Iván; Ibarra, Angel

    2017-07-01

    The research study focuses on the neutronic design analysis and optimization of one of the options for a fusion reactor designed as DCLL (dual coolant lithium-lead). The main objective has been to develop an efficient and technologically viable modular DCLL breeding blanket (BB) using the DEMO generic design specifications established within the EUROfusion Programme. The final neutronic design has to satisfy the requirements of: tritium self-sufficiency; BB thermal efficiency; preservation of plasma confinement; temperature limits imposed by materials; and radiation limits to guarantee the largest operational life for all the components. Therefore, a 3D fully heterogeneous DCLL neutronic model has been developed for the DEMO baseline 2014 determining its behaviour under the real operational conditions of the DEMO reactor. Consequent actions have been adopted to improve its performances. Neutronic assessments have specially addressed tritium breeding ratio, multiplication energy factor, power density distributions, damage and shielding responses. The model has then been adapted to the subsequent DEMO baseline 2015 (with a more powerful and bigger plasma, smaller divertor and bigger blanket segments), implying new design choices to improve the reactor nuclear performances.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  9. Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor *

    Directory of Open Access Journals (Sweden)

    Parma Edward J.

    2016-01-01

    Full Text Available Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity “bucket” environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters.

  10. Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor

    Science.gov (United States)

    Parma, Edward J.; Naranjo, Gerald E.; Lippert, Lance L.; Vehar, David W.

    2016-02-01

    Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR) is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity "bucket" environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters. Work supported by the United States Department of Energy at Sandia National Laboratories. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned

  11. Study on operation of a research reactor during one PCS pump failure accident

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Kyoung Woo; Yoon, Hyu Ngi; Kim, Seong Hoon; Chi, Dae Young; Yoon, Juh Yeon [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    The Primary cooling system (PCS) of a research reactor is designed to provide adequate cooling to the reactor core with a reasonable margin during all operation modes. The PCS consists of pumps, heat exchangers, and all necessary interconnecting pipes, valves, and instruments. The number of pumps is determined from a safety and economic point of view. As the number of pump trains increase, the cost increases according to the increase in safety class equipment. However, it is impossible to install one pump for a PCS because a zero flow can instantaneously occur during a pump failure such as a pump seizure. Thus, a PCS frequently consists of two parallel 50% capacity pumps and heat exchangers. In addition, check valves are generally installed to prevent a reversal flow when multiple pumps are designed to operate. However, if a swing type check valve is used, it should be estimated whether the slam due to instantaneous closing of the valve affects the system vibration. To reduce the vibration by a slam phenomenon, additional equipment such as a damper will be installed in the valve. The purpose of the check valve in PCS is to prevent the flow path when a reverse flow occurs. The installation of additional equipment will make it difficult to perform this function. In this study, it is estimated whether the PCS can operate without check valves. First, a flow analysis using Flowmaster was compared and verified by the calculation employing a empirical correlation. Second, the simulation for a one pump failure accident was performed and analyzed.

  12. Research and development in welding and hardfacing towards construction of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Albert, S.K.; Bhaduri, A.K. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2012-07-01

    India's 500MWe Prototype Fast Breeder Reactor (PFBR) is in advanced stage of construction at Kalpakkam and this reactor is expected to be commissioned in the year 2013. Extensive research and development activities in various fields like material development, welding, forming, non-destructive testing etc. were undertaken before the actual construction of the reactor began. Many of these activities are still continuing with the objectives of conducting functional tests, generating data, validating the design and meeting the various regulatory requirements. In welding, initial challenge was to develop indigenous welding consumables with a specification more stringent than that is given in most of the national and international standards. The welding consumable specified for 316LN austenitic stainless steel is E316-15M with strict control on delta ferrite content, toughness requirement after 750 C/100 h ageing to ensure adequate resistance to embrittlement during prolonged high temperature exposure and good slag detachability. This consumable was successfully developed in collaboration with Indian consumable manufacturers and is being used for fabrication of almost all PFBR components and piping made of 316LN stainless steel. Similarly, electrodes of welding of modified 9Cr-1Mo steel (material of construction for PFBR steam generator) with requirement of RTNDT requirement of ≤ -5 C was also developed indigenously. Extensive studies were also carried out on weldability of various austenitic stainless steels and modified 9Cr-1Mo steel used in PFBR. Hot cracking susceptibility of alloy D9 (15Cr-15Ni-2Mo-Ti alloy), the material chosen for fuel clab tube and fuel sub-assembly, and 316LN stainless steel was extensively studied using varestraint testing and Gleeble simulation. Results from these steels were used in developing welding procedures for various reactor components. Hydrogen assisted cracking susceptibility (HAC) of modified 9Cr-1Mo steel was studied using

  13. Theoretical Study of Palladium Membrane Reactor Performance During Propane Dehydrogenation Using CFD Method

    Directory of Open Access Journals (Sweden)

    Kamran Ghasemzadeh

    2017-04-01

    Full Text Available This study presents a 2D-axisymmetric computational fluid dynamic (CFD model to investigate the performance Pd membrane reactor (MR during propane dehydrogenation process for hydrogen production. The proposed CFD model provided the local information of temperature and component concentration for the driving force analysis. After investigation of mesh independency of CFD model, the validation of CFD model results was carried out by other modeling data and a good agreement between CFD model results and theoretical data was achieved. Indeed, in the present model, a tubular reactor with length of 150 mm was considered, in which the Pt-Sn-K/Al2O3 as catalyst were filled in reaction zone. Hence, the effects of the important operating parameter (reaction temperature on the performances of membrane reactor (MR were studied in terms of propane conversion and hydrogen yield. The CFD results showed that the suggested MR system during propane dehydrogenation reaction presents higher performance with respect to once obtained in the conventional reactor (CR. In particular, by applying Pd membrane, was found that propane conversion can be increased from 41% to 49%. Moreover, the highest value of propane conversion (X = 91% was reached in case of Pd-Ag MR. It was also established that the feed flow rate of the MR is to be the one of the most important factors defining efficiency of the propane dehydrogenation process.

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

    Science.gov (United States)

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

    2003-01-01

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

  15. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  16. Performance Test of System Identification Methods for a Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Keuk Jong; Kim, Han Gon [KHNP, Daejeon (Korea, Republic of)

    2011-05-15

    An automatic controller that uses the model predictive control (MPC) method is being developed for automatic load follow operation. As described in Ref. a system identification method is important in the MPC method because MPC is based on a system model produced by system identification. There are many models and methods of system identification. In this study, AutoRegressive eXogenous (ARX) model was selected from among them, and the recursive least square (RLS) method and least square (LS) method associated with this model are used in a comparative performance analysis

  17. Effect of Different Structural Materials on Neutronic Performance of a Hybrid Reactor

    Science.gov (United States)

    Übeyli, Mustafa; Tel, Eyyüp

    2003-06-01

    Selection of structural material for a fusion-fission (hybrid) reactor is very important by taking into account of neutronic performance of the blanket. Refractory metals and alloys have much higher operating temperatures and neutron wall load (NWL) capabilities than low activation materials (ferritic/martensitic steels, vanadium alloys and SiC/SiC composites) and austenitic stainless steels. In this study, effect of primary candidate refractory alloys, namely, W-5Re, T111, TZM and Nb-1Zr on neutronic performance of the hybrid reactor was investigated. Neutron transport calculations were conducted with the help of SCALE 4.3 System by solving the Boltzmann transport equation with code XSDRNPM. Among the investigated structural materials, tantalum had the worst performance due to the fact that it has higher neutron absorption cross section than others. And W-5Re and TZM having similar results showed the best performance.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-01

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

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

    Directory of Open Access Journals (Sweden)

    Kaiser Krista

    2016-01-01

    Full Text Available 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-10-06

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-31

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

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

    Science.gov (United States)

    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.

  6. International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

    2014-12-15

    Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

  7. Improvement of the reactivity computer for HANARO research reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-04-01

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

  8. Performance of static var compensator control type thyristor controlled reactor and thyristor switched capacitor

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Josias M. de; Yung, Chou Shaw; Rose, Eber H.; Pantoja, Antonio L.A. [ELETRONORTE, Belem, PA (Brazil); Fouesnant, Thomas; Boissier, Luc

    1994-12-31

    This paper has the objective of presenting the philosophy of Static Var Compensator (SVC) Control as well the necessary adjustments in the project of control system to guarantee suitable performance under different operating conditions. The verification on the performance of the SVC control has been done by Transient Network Analyzer (TNA/CEPEL) studies, commissioning tests and a factory tests. The SVC is the type of Thyristor Controlled Reactor (TCR) and Thyristor Switched Capacitor (TSC). (author) 3 refs., 12 figs.

  9. The upgrade and conversion of the ET-RR-1 research reactor using plate type fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Ashoub, N. [Reactor Physics Dept., Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt); Saleh, H.G. [Faculty of Girls for Arts and Education, Ain-Shams Univ., Cairo (Egypt)

    2001-11-01

    The ET-RR-1 research reactor has been operated at 2 MW since 1961 using EK-10 fuel elements with 10% enriched uranium. The reactor has been used for nuclear applied research and isotope production. In order to upgrade the reactor power to a reasonable limit facing up-to-date uses, core conversion by a new type of fuel element available is necessary. Two fuel elements in plate type are suggested in this study to be used in the ET-RR-1 reactor core rather than the utilized ones. The first element has a dimension of 8 x 8 x 50 cm and consists of 19.7% enriched uranium, which is typical for that utilized in the ET-RR-2 reactor, but with a different length. The other element is proposed with a dimension of 7 x 7 x 50 cm and has the same uranium enrichment. To accomplish safety requirements for these fuel elements, thermal-hydraulic evaluation has been carried out using the PARET code. To reach a core conversion of the ET-RR-1 reactor with the above two types of fuel elements, neutronic calculations have been performed using WIMSD4, DIXY2 and EREBUS codes. Some important nuclear parameters needed in the physical design of the reactor were calculated and included in this study. (orig.) [German] Der ET-RR-1 Forschungsreaktor wird seit 1961 unter Verwendung von EK-10 Brennelementen mit einer Leistung von 2 MW betrieben. Der Reaktor wird in der angewandten Forschung und zur Isotopenherstellung eingesetzt. Um die Reaktorleistung im Hinblick auf eine zeitgemaesse Nutzung der Anlage in einem vernuenftigen Mass zu erhoehen, ist eine Umwandlung des Kerns durch Verwendung neuartiger Brennelemente noetig. In der vorliegenden Untersuchung wird vorgeschlagen, anstelle der z. Z. verwendeten Elemente zwei neue, plattenfoermige Brennelemente zu verwenden. Das erste Element hat eine Groesse von 8 x 8 x 50 cm und besteht aus 19,7% angereichertem Uran, was den im ET-RR-2 Reaktor verwendeten Elementen entspricht, allerdings mit einer anderen Groesse. Das zweite Element hat die gleiche

  10. Reactors licensing: proposal of an integrated quality and environment regulatory structure for nuclear research reactors in Brazil; Licenciamento de reatores: proposta de uma estrutura regulatoria integrada com abordagem em qualidade e meio ambiente para reatores de pesquisa no Brasil

    Energy Technology Data Exchange (ETDEWEB)

    Serra, Reynaldo Cavalcanti

    2014-07-01

    A new integrated regulatory structure based on quality and integrated issues has been proposed to be implemented on the licensing process of nuclear research reactors in Brazil. The study starts with a literature review about the licensing process in several countries, all of them members of the International Atomic Energy Agency. After this phase it is performed a comparative study with the Brazilian licensing process to identify good practices (positive aspects), the gaps on it and to propose an approach of an integrated quality and environmental management system, in order to contribute with a new licensing process scheme in Brazil. The literature review considered the following research nuclear reactors: Jules-Horowitz and OSIRIS (France), Hanaro (Korea), Maples 1 and 2 (Canada), OPAL (Australia), Pallas (Holand), ETRR-2 (Egypt) and IEA-R1 (Brazil). The current nuclear research reactors licensing process in Brazil is conducted by two regulatory bodies: the Brazilian National Nuclear Energy Commission (CNEN) and the Brazilian Institute of Environment and Renewable Natural Resources (IBAMA). CNEN is responsible by nuclear issues, while IBAMA by environmental one. To support the study it was applied a questionnaire and interviews based on the current regulatory structure to four nuclear research reactors in Brazil. Nowadays, the nuclear research reactor’s licensing process, in Brazil, has six phases and the environmental licensing process has three phases. A correlation study among these phases leads to a proposal of a new quality and environmental integrated licensing structure with four harmonized phases, hence reducing potential delays in this process. (author)

  11. The Denitrifying Biological Phosphorus Removal Performance in Anaerobic/Anoxic Sequencing Batch Reactor: The Effect of Carbon Source

    OpenAIRE

    Gürtekin, Engin; ŞEKERDAĞ, Nusret

    2015-01-01

    In this study, the effect of carbon source on denitrifying biological phosphorus removal performance in acetate and glucose fed two anaerobic/anoxic sequencinq batch reactor (SBR) was investigated. Glucose and acetate were used as the substrates. In acetate and glucose fed reactors, the COD (Chemical Oxygen Demand) removal efficiencies were 91,90% and PO4-P removal efficiencies were 87,51% respectively. These results shows that the phosphorus removal efficiency is lower in glucose fed reactor.

  12. The Denitrifying Biological Phosphorus Removal Performance in Anaerobic/Anoxic Sequencing Batch Reactor: The Effect of Carbon Source

    OpenAIRE

    Gürtekin, Engin; ŞEKERDAĞ, Nusret

    2015-01-01

    In this study, the effect of carbon source on denitrifying biological phosphorus removal performance in acetate and glucose fed two anaerobic/anoxic sequencinq batch reactor (SBR) was investigated. Glucose and acetate were used as the substrates. In acetate and glucose fed reactors, the COD (Chemical Oxygen Demand) removal efficiencies were 91,90% and PO4-P removal efficiencies were 87,51% respectively. These results shows that the phosphorus removal efficiency is lower in glucose fed reactor.

  13. Improvement of Membrane Performances to Enhance the Yield of Vanillin in a Pervaporation Reactor

    Directory of Open Access Journals (Sweden)

    Giovanni Camera-Roda

    2014-02-01

    Full Text Available In membrane reactors, the interaction of reaction and membrane separation can be exploited to achieve a “process intensification”, a key objective of sustainable development. In the present work, the properties that the membrane must have to obtain this result in a pervaporation reactor are analyzed and discussed. Then, the methods to enhance these properties are investigated for the photocatalytic synthesis of vanillin, which represents a case where the recovery from the reactor of vanillin by means of pervaporation while it is produced allows a substantial improvement of the yield, since its further oxidation is thus prevented. To this end, the phenomena that control the permeation of both vanillin and the reactant (ferulic acid are analyzed, since they ultimately affect the performances of the membrane reactor. The results show that diffusion of the aromatic compounds takes place in the presence of low concentration gradients, so that the process is controlled by other phenomena, in particular by the equilibrium with the vapor at the membrane-permeate interface. On this basis, it is demonstrated that the performances are enhanced by increasing the membrane thickness and/or the temperature, whereas the pH begins to limit the process only at values higher than 6.5.

  14. Improvement of membrane performances to enhance the yield of vanillin in a pervaporation reactor.

    Science.gov (United States)

    Camera-Roda, Giovanni; Cardillo, Antonio; Loddo, Vittorio; Palmisano, Leonardo; Parrino, Francesco

    2014-02-28

    In membrane reactors, the interaction of reaction and membrane separation can be exploited to achieve a "process intensification", a key objective of sustainable development. In the present work, the properties that the membrane must have to obtain this result in a pervaporation reactor are analyzed and discussed. Then, the methods to enhance these properties are investigated for the photocatalytic synthesis of vanillin, which represents a case where the recovery from the reactor of vanillin by means of pervaporation while it is produced allows a substantial improvement of the yield, since its further oxidation is thus prevented. To this end, the phenomena that control the permeation of both vanillin and the reactant (ferulic acid) are analyzed, since they ultimately affect the performances of the membrane reactor. The results show that diffusion of the aromatic compounds takes place in the presence of low concentration gradients, so that the process is controlled by other phenomena, in particular by the equilibrium with the vapor at the membrane-permeate interface. On this basis, it is demonstrated that the performances are enhanced by increasing the membrane thickness and/or the temperature, whereas the pH begins to limit the process only at values higher than 6.5.

  15. Performance of low smeared density sodium-cooled fast reactor metal fuel

    Science.gov (United States)

    Porter, D. L.; Chichester, H. J. M.; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-10-01

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-08-15

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

  17. The power distribution and neutron fluence measurements and calculations in the VVER-1000 Mock-Up on the LR-0 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kostal, M.; Juricek, V.; Rypar, V.; Svadlenkova, M. [Research Center Rez Ltd., 250 68 Husinec-Rez 130 (Czech Republic); Cvachovec, F. [Univ. of Defence, Kounicova 65, 662 10 Brno (Czech Republic)

    2011-07-01

    The power density distribution in a reactor has significant influence on core structures and pressure vessel mechanical resistance, as well as on the physical characteristics of nuclear fuel. This quantity also has an effect on the leakage neutron and photon field. This issue has become of increasing importance, as it touches on actual questions of the VVER nuclear power plant life time extension. This paper shows the comparison of calculated and experimentally determined pin by pin power distributions. The calculation has been performed with deterministic and Monte Carlo approaches. This quantity is accompanied by the neutron and photon flux density calculation and measurements at different points of the light water zero-power (LR-0) research reactor mock-up core, reactor built-in component (core barrel), and reactor pressure vessel and model. The effect of the different data libraries used for calculation is discussed. (authors)

  18. Physicochemical and Microbial Caracteristics Performency in Wastewater Treated Under Aerobic Reactor

    Directory of Open Access Journals (Sweden)

    Asma B. Rajeb

    2011-01-01

    Full Text Available Problem statement: The current work study the efficiency of biological wastewater treatment by an aerobic reactor which could be used in small agglomerations. RBC reduced physicochemical and microbiological load of wastewater but values remain above Tunisian standard. Approach: Experiments were conducted on a sand filled PVC column fed with wastewater treated by Rotating Biological Contactor (RBC at a pulsed rhythm of 8 sequences per day. For performances study process, physicochemical and bacterial analyses effluent at inlet and outlet of column were realized. Results: The results showed that through filter mass (D10 = 0.55 mm, D60 = 1.3 mm and coefficient uniformity = 2.36 96% of suspended solids, 99% of NH4 +-N (during first phase, 92% of COD, 91% of BOD5 and 46% of phosphorus are retained by surface filtration. The microbial abatement results is E. coli. The microbial water quality is slightly higher than Tunisian standards. The removal of microbial indicators in the considered reactor depends on the depth of the filter and negatively correlated with NO3 --N (r = -0.99, with E. coli at 3rd OPD. Conclusion/Recommendation: Results confirmed that the reactor tested is performed as an advanced treatment system for DBO, COD, SS, NH4 +-N and NO3 --N. Despite that 96% of SS efficiency reduction, clogging is not achieved quickly that due to biofilm detachment phenomena. The removal of microbial indicators in the considered reactor depends on the depth of the filter and negatively correlated with NO3 --N. Disinfection performances for the considered reactor reduce microbial load, however chlore, ozone or UV disinfection should be considered.

  19. Irradiation performance of AGR-1 high temperature reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; John D. Hunn; Robert N. Morris; Charles A. Baldwin; Philip L. Winston; Jason M. Harp; Scott A. Ploger; Tyler Gerczak; Isabella J. van Rooyen; Fred C. Montgomery; Chinthaka M. Silva

    2014-10-01

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO-coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.5% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel–including the extent of fission product release and the evolution of kernel and coating microstructures–was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocrabon and compact matrix. The capsule-average fractional release from the compacts was 1×10 4 to 5×10 4 for 154Eu and 8×10 7 to 3×10 5 for 90Sr. The average 134Cs release from compacts was <3×10 6 when all particles maintained intact SiC. An estimated four particles out of 2.98×105 experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs release in two capsules to approximately 10 5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. Palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization

  20. 77 FR 26321 - Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112

    Science.gov (United States)

    2012-05-03

    ... COMMISSION Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112 AGENCY... License No. R- 112, held by Reed College (the licensee), which authorizes continued operation of the Reed... renewed Facility Operating License No. R-112 will expire 20 years from its date of issuance. The...

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

    DEFF Research Database (Denmark)

    Goodstein, L.P.

    1969-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Head, C.R.

    1997-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-02-01

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

  5. Neutron fluence depth profiles in water phantom on epithermal beam of LVR-15 research reactor.

    Science.gov (United States)

    Viererbl, L; Klupak, V; Lahodova, Z; Marek, M; Burian, J

    2010-01-01

    Horizontal channel with epithermal neutron beam at the LVR-15 research reactor is used mainly for boron neutron capture therapy. Neutron fluence depth profiles in a water phantom characterise beam properties. The neutron fluence (approximated by reaction rates) depth profiles were measured with six different types of activation detectors. The profiles were determined for thermal, epithermal and fast neutrons.

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

    Science.gov (United States)

    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.

  7. 75 FR 62892 - Massachusetts Institute of Technology Research Reactor Environmental Assessment and Finding of No...

    Science.gov (United States)

    2010-10-13

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

  8. A Study on Comparison of HANARO and KIJANG Research Reactor in Nuclear Safeguards

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Juang; Lee, Sung Ho; Kim, Hyun-Jo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    As one of major national projects for nuclear science and engineering in Korea, the KIJANG Research Reactor(KJRR) project was commenced in order to develop the core research reactor(RR) technologies for strengthening the competitiveness of the RR export and also to stabilize the supply of key radioisotopes for medical and industrial applications. This paper is about applying IAEA safeguards at new nuclear facility (KJRR). The beginning of this project is comparing of HANARO and KIJANG research reactor in nuclear safeguards for nuclear material accountancy method. As mentioned before, research reactor is basically item counting facility. In Fig 1, first two processes are belonging to item counting. But last two processes are for bulk handling. So KIJANG RR would be treated item counting facility as well as bulk handling facility by fission moly production facility. For this reason, nuclear material accountancy method for KJRR is not easy compared to existing one. This paper accounted for solution of KJRR nuclear material accountancy briefly. Future study on the suitable nuclear material accountancy method for mixed facility between item counting facility and bulk handling facility will be conducted more specifically.

  9. Comparative performance of fixed-film biological filters: Application of reactor theory

    Science.gov (United States)

    Watten, B.J.; Sibrell, P.L.

    2006-01-01

    Nitrification is classified as a two-step consecutive reaction where R1 represents the rate of formation of the intermediate product NO2-N and R2 represents the rate of formation of the final product NO3-N. The relative rates of R1 and R2 are influenced by reactor type characterized hydraulically as plug-flow, plug-flow with dispersion and mixed-flow. We develop substrate conversion models for fixed-film biofilters operating in the first-order kinetic regime based on application of chemical reactor theory. Reactor type, inlet conditions and the biofilm kinetic constants Ki (h-1) are used to predict changes in NH4-N, NO2-N, NO3-N and BOD5. The inhibiting effects of the latter on R1 and R2 were established based on the ?? relation, e.g.:{A formula is presented}where BOD5,max is the concentration that causes nitrification to cease and N is a variable relating Ki to increasing BOD5. Conversion models were incorporated in spreadsheet programs that provided steady-state concentrations of nitrogen and BOD5 at several points in a recirculating aquaculture system operating with input values for fish feed rate, reactor volume, microscreen performance, make-up and recirculating flow rates. When rate constants are standardized, spreadsheet use demonstrates plug-flow reactors provide higher rates of R1 and R2 than mixed-flow reactors thereby reducing volume requirements for target concentrations of NH4-N and NO2-N. The benefit provided by the plug-flow reactor varies with hydraulic residence time t as well as the effective vessel dispersion number, D/??L. Both reactor types are capable of providing net increases in NO2-N during treatment but the rate of decrease in the mixed-flow case falls well behind that predicted for plug-flow operation. We show the potential for a positive net change in NO2-N increases with decreases in the dimensionless ratios K2, (R2 )/K1,( R1 ) and [NO2-N]/[NH4-N] and when the product K1, (R1) t provides low to moderate NH4-N conversions. Maintaining

  10. Neutron spectrum measurements at a radial beam port of the NUR research reactor using a Bonner spheres spectrometer.

    Science.gov (United States)

    Mazrou, H; Nedjar, A; Seguini, T

    2016-08-01

    This paper describes the measurement campaign held around the neutron radiography (NR) facility of the Algerian 1MW NUR research reactor. The main objective of this work is to characterize accurately the neutron beam provided at one of the radial channels of the NUR research reactor taking benefit of the acquired CRNA Bonner spheres spectrometer (BSS). The specific objective was to improve the image quality of the NR facility. The spectrometric system in use is based on a central spherical (3)He thermal neutron proportional counter combined with high density polyethylene spheres of different diameters ranging from 3 to 12in. This counting system has good gamma ray discrimination and is able to cover an energy range from thermal to 20MeV. The measurements were performed at the sample distance of 0.6m from the beam port and at a height of 1.2m from the facility floor. During the BSS measurements, the reactor was operating at low power (100W) to avoid large dead times, pulse pileup and high level radiation exposures, in particular, during spheres handling. Thereafter, the neutron spectrum at the sample position was unfolded by means of GRAVEL and MAXED computer codes. The thermal, epithermal and fast neutron fluxes, the total neutron flux, the mean energy and the Cadmium ratio (RCd) were provided. A sensitivity analysis was performed taking into account various defaults spectra and ultimately a different response functions in the unfolding procedure. Overall, from the obtained results it reveals, unexpectedly, that the measured neutron spectrum at the sample position of the neutron radiography of the NUR reactor is being harder with a predominance of fast neutrons (>100keV) by about 60%. Finally, those results were compared to previous and more recent measurements obtained by activation foils detectors. The agreement was fairly good highlighting thereby the consistency of our findings.

  11. Application and facility of neutron activation analysis in HANARO research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y.S. [Korea Atomic Energy Research Institute, Taejon (Korea)

    2001-11-01

    The facilities for neutron activation analysis in the HANARO research reactor are described and the main applications of NAA in Korea are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system, were installed at three irradiation holes at the end of 1995. One irradiation hole is lined with a cadmium tube for epithermal NAA. The performance of the NAA facility was examined to identify the characteristics of tube transfer system, irradiation sites and custom made polythylene irradiation capsule. The available thermal neutron flux with each irradiation site are in the range of 3 x 10{sup 13} {approx} 1 x 10{sup 14} n/cm{sup 2}{center_dot}s and cadmium ratios are 15 {approx} 250. For an automatic sample changer for gamma-ray counting, a domestic product was designed and manufactured. An integrated computer program (Labview) for the calculation of content was developed. Neutron activation analysis has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials and various polymers for research and development. Improvement of analytical procedures and establishment of an analytical quality control and assurance system were studied. Applied research and development for the environment, industry and human health by NAA and its standardization was carried out. For the application of the Korea Laboratory Accreditation Scheme (KOLAS), evaluation of measurement uncertainty and proficiency testing of reference materials was performed. Also to verify the reliability and validity of analytical results, intercomparison studies between laboratories were carried out. In this paper, analytical services, national cooperation and the results of the researches are summarized. (author)

  12. 77 FR 13376 - Notice of License Termination for the University of Arizona Research Reactor, License No. R-52

    Science.gov (United States)

    2012-03-06

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

  13. 77 FR 4807 - Revised Fee Policy for Acceptance of Foreign Research Reactor Spent Nuclear Fuel From High-Income...

    Science.gov (United States)

    2012-01-31

    ... National Nuclear Security Administration Revised Fee Policy for Acceptance of Foreign Research Reactor... Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel'' (61 FR 25092, May..., Department of Energy. ACTION: Notice of a change in the fee policy. SUMMARY: This notice announces a...

  14. Optimization of the core configuration design using a hybrid artificial intelligence algorithm for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hedayat, Afshin, E-mail: ahedayat@aut.ac.i [Department of Nuclear Engineering and Physics, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, P.O. Box 15875-4413, Tehran (Iran, Islamic Republic of); Reactor Research and Development School, Nuclear Science and Technology Research Institute (NSTRI), End of North Karegar Street, P.O. Box 14395-836, Tehran (Iran, Islamic Republic of); Davilu, Hadi [Department of Nuclear Engineering and Physics, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, P.O. Box 15875-4413, Tehran (Iran, Islamic Republic of); Barfrosh, Ahmad Abdollahzadeh [Department of Computer Engineering, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, P.O. Box 15875-4413, Tehran (Iran, Islamic Republic of); Sepanloo, Kamran [Reactor Research and Development School, Nuclear Science and Technology Research Institute (NSTRI), End of North Karegar Street, P.O. Box 14395-836, Tehran (Iran, Islamic Republic of)

    2009-12-15

    To successfully carry out material irradiation experiments and radioisotope productions, a high thermal neutron flux at irradiation box over a desired life time of a core configuration is needed. On the other hand, reactor safety and operational constraints must be preserved during core configuration selection. Two main objectives and two safety and operational constraints are suggested to optimize reactor core configuration design. Suggested parameters and conditions are considered as two separate fitness functions composed of two main objectives and two penalty functions. This is a constrained and combinatorial type of a multi-objective optimization problem. In this paper, a fast and effective hybrid artificial intelligence algorithm is introduced and developed to reach a Pareto optimal set. The hybrid algorithm is composed of a fast and elitist multi-objective genetic algorithm (GA) and a fast fitness function evaluating system based on the cascade feed forward artificial neural networks (ANNs). A specific GA representation of core configuration and also special GA operators are introduced and used to overcome the combinatorial constraints of this optimization problem. A software package (Core Pattern Calculator 1) is developed to prepare and reform required data for ANNs training and also to revise the optimization results. Some practical test parameters and conditions are suggested to adjust main parameters of the hybrid algorithm. Results show that introduced ANNs can be trained and estimate selected core parameters of a research reactor very quickly. It improves effectively optimization process. Final optimization results show that a uniform and dense diversity of Pareto fronts are gained over a wide range of fitness function values. To take a more careful selection of Pareto optimal solutions, a revision system is introduced and used. The revision of gained Pareto optimal set is performed by using developed software package. Also some secondary operational

  15. Fuel irradiation research of Japan at halden reactor. Achievement of cooperative researches between JAERI and several organizations in the period from 1997 to 1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-11-01

    JAERI has performed cooperative researches with several Japanese agencies and companies by means of the Halden Boiling Heavy Water Reactor (HBWR) which is located at Halden in Norway. These researches are carried out based on the contracts of the cooperative researches, which are revised every three years, in accordance with the renewal of the participation of JAERI to the OECD Halden Reactor Project. This report summaries the objectives, contents and the outlines of the achievements of the cooperative researches during the three years from 1997 January to 1999 December. During the period, nine cooperative researches had been carried out. Two of them had been completed and other seven researches has been continued to the next three years period. Most of them are irradiation test researches of advanced fuel and cladding in order to prepare the higher burnup utilization and introduction of LWR fuel and MOX fuel in LWRs of Japan. The researches of fuel irradiation usually take long time for preparing test and irradiation, then three years are usually not enough to obtain some achievements from the irradiation tests. Therefore, seven tests have been continued to the next three year period. In this report, the achievements of the continued researches to the next period are not final one but a kind of progress report. (author)

  16. Characterization of a New Continuous Air Monitoring System For the University of Massachusetts Lowell Research Reactor

    Science.gov (United States)

    Alqahtani, Mohammad Saad

    A continuous air monitor (CAM) is a critical piece of equipment to support radiation safety in nuclear facilities where the generation of airborne radioactivity is a possibility for either normal operations or accident scenarios. The University of Massachusetts Lowell Research Reactor is planning to install a new CAM system manufactured by Canberra Industries for monitoring airborne radioactive particulates. In this study, the new CAM was evaluated to determine 1) baseline response, 2) response to high exposure rates, 3) appropriate background compensation, 4) detection limits, and 5) alarm settings. The results of this study will help to properly integrate the new CAM into the reactor radiation monitoring system.

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seung Kook; Chung, K.H

    1999-04-01

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

  18. Fuel burnup analysis for Thai research reactor by using MCNPX computer code

    Science.gov (United States)

    Sangkaew, S.; Angwongtrakool, T.; Srimok, B.

    2017-06-01

    This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.

  19. Recent expansion of research for light water reactor technology

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Kiichi (Japan Electric Power Information Center, Tokyo (Japan)); Nemoto, Kazuyasu; Aoki, Norichika; Kusanagi, Hideo

    1990-09-01

    It is needless to say that for simultaneously coping with the increase of energy consumption and the prevention of the worsening of environment in the world, and for maintaining the standard of living in Japan where energy resources are scarce, the development of atomic energy is necessary. Though the technology of LWRs has been already established, the efforts of aiming at the further high safety and reliability of LWRs must be exerted. In this report, the recent technical development is described, centering around the research and technical development promoted by the Central Research Institute of Electric Power Industry. The energy consumption in the world recorded the yearly growth of about 3%, and in 1987, it was 9.65 billion tons in terms of coal (7000 kcal/kg). The problems of earth environment will relax by promoting atomic energy. As for the recent development of LWR technology, the research on existing LWRs, the research on the LWRs of next generation, the research on the new technology for locating nuclear facilities and the research on radiation are carried out. As the research aiming at the LWRs of next generation, the design and evaluation of statically safe LWRs, the evaluation of fuel behavior at high burnup and the development of new location technology are carried out. (K.I.).

  20. Preliminary Assessment of a Debris Bed Cooling Performance for Demonstration Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Chung Ho; Park, Chang Gyu; Song, Hoon; Kim, Young Gyun; Jeong, Hae Yong; Chang, Jin Wook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    In the case of the sodium-cooled fast reactor such as KALIMER-600, Hypothetical Core Disruptive Accident (HCDA) attributed from mass nuclear fuel melting is unlikely to occur due to defense in depth concepts to meet requirements of redundancy and diversity. Multiple faults such as loss of flow, loss of heat sink, or transient overpower without scram are to lead rising the power level until cladding failure as reactivity increasing. The fact that metallic fuel melts at a lower temperature than the cladding allows significant in-pin- fuel motion to occur prior to cladding failure. Also, the combination of Doppler and axial expansion feedback and negative feedback associated with the in-pin fuel relocation prevents the reactivity from reaching prompt critical. Finally, the resulting reactivity and power reductions help prevent fuel temperatures from rising more than the fuel melting temperature. It is more difficult to occur HCDA in a metallic fueled core because reactor power and heat removal capability is maintained in balance by inherent safety characteristics However, for the future design of sodium-cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth considering due to the triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. Accordingly, evaluation of a packed debris bed cooling performance with single phase flow for demonstration sodium-cooled fast reactor was carried out for proof of the in-vessel retention of the core debris

  1. Bright Flash Neutron Radiography at the McClellan Nuclear Research Reactor

    Science.gov (United States)

    Lerche, M.; Tremsin, A. S.; Schillinger, B.

    The University of California, Davis McClellan Nuclear Research Center (MNRC) operates a 2 MW TRIGATM reactor, which is currently the highest power TRIGATM reactor in the United States. The Center was originally build by the US Air Force to detect hidden defects in aircraft structures using neutron radiography; the Center can accommodate samples as large as 10.00 m long, 3.65 m high, and weighing up to 2,270 kg. The MNRC reactor can be pulsed to 350 MW for about 30 ms (FWHM). The combination of a short neutron pulse with a fast microchannel plate based neutron detector enables high-resolution flash neutron radiography to complement conventional neutron radiography

  2. Disposal of beryllium and cadmium from research reactors; Entsorgung von Beryllium und Cadmium aus Forschungsreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Lierse von Gostomski, C.; Remmert, A.; Stoewer, W. [Inst. fuer Radiochemie, Technische Univ. Muenchen, Garching (Germany); Bach, F.W.; Wilk, P.; Kutlu, I. [Inst. fuer Werkstoffkunde, Univ. Hannover, Hannover (Germany); Blenski, H.J.; Berthold, M. [Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Nerlich, K.D.; Plank, W. [TUeV Sueddeutschland Bau und Betrieb GmbH, Muenchen (Germany)

    2003-07-01

    Beryllium and cadmium mostly occur in metal form as radioactive special materials during the deconstruction of research reactors. Beryllium is usually used in these reactors as a neutron reflector and moderator, while cadmium is used above all as a neutron absorber. Both metals together have a high chemotoxicity as well as an inventory of radionuclides which has not been more closely characterised up to now. A high tritium content is to be expected, particularly in the case of beryllium; this tritium is due to the reaction of the metal with thermal reactor neutrons in particular. However, other nuclides which may be formed by neutron capture from impurities also contribute to the activity inventory. Up to now there is no qualified process for proper treatment, conditioning and intermediate and final repository in Germany. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-12-01

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

  5. Preliminary conceptual design for electrical and I and C system of a new research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, Y. K.; Kim, M. J.; Kim, H. K.; Ryu, J. S

    2004-01-01

    The core type and the process system design will be varied according to the reactor's application and capacity. A New research reactor is being designed by KAERI since 2002 and the process systems are not fixed yet. But control and instrument systems are similar to each other even though the application and the size are not same. So the C and I system that encompasses reactor protection system, reactor control system, and computer system was designed conceptually according to the requirements based on new digital technology and HANARO's proven design. The plant electrical system consists of off-site system that delivers bulk electrical power to the reactor site and on-site system that distributes and controls electrical power at the facility. The electrical system includes building service system that consist of lighting, communication, fire detection, grounding, cathodic protection, etc. also. This report describes the design requirements of on-site and off-site electric power system that set up from the codes and standards and the conceptual design based on the design requirements.

  6. Effect of post-digestion temperature on serial CSTR biogas reactor performance.

    Science.gov (United States)

    Boe, Kanokwan; Karakashev, Dimitar; Trably, Eric; Angelidaki, Irini

    2009-02-01

    The effect of post-digestion temperature on a lab-scale serial continuous-flow stirred tank reactor (CSTR) system performance was investigated. The system consisted of a main reactor operated at 55 degrees C with hydraulic retention time (HRT) of 15 days followed by post-digestion reactors with HRT of 5.3 days. Three post-digestion temperatures (55 degrees C, 37 degrees C and 15 degrees C) were compared in terms of biogas production, process stability, microbial community and methanogenic activity. The results showed that the post-digesters operated at 55 degrees C, 37 degrees C and 15 degrees C gave extra biogas production of 11.7%, 8.4% and 1.2%, respectively. The post-digester operated at 55 degrees C had the highest biogas production and was the most stable in terms of low VFA concentrations. The specific methanogenic activity tests revealed that the main reactor and the post-digester operated at 55 degrees C and 37 degrees C had very active acidogens and methanogens. In contrast, very low methanogenic activity was observed at 15 degrees C.

  7. Pressure-accelerated azide-alkyne cycloaddition: micro capillary versus autoclave reactor performance.

    Science.gov (United States)

    Borukhova, Svetlana; Seeger, Andreas D; Noël, Timothy; Wang, Qi; Busch, Markus; Hessel, Volker

    2015-02-01

    Pressure effects on regioselectivity and yield of cycloaddition reactions have been shown to exist. Nevertheless, high pressure synthetic applications with subsequent benefits in the production of natural products are limited by the general availability of the equipment. In addition, the virtues and limitations of microflow equipment under standard conditions are well established. Herein, we apply novel-process-window (NPWs) principles, such as intensification of intrinsic kinetics of a reaction using high temperature, pressure, and concentration, on azide-alkyne cycloaddition towards synthesis of Rufinamide precursor. We applied three main activation methods (i.e., uncatalyzed batch, uncatalyzed flow, and catalyzed flow) on uncatalyzed and catalyzed azide-alkyne cycloaddition. We compare the performance of two reactors, a specialized autoclave batch reactor for high-pressure operation up to 1800 bar and a capillary flow reactor (up to 400 bar). A differentiated and comprehensive picture is given for the two reactors and the three methods of activation. Reaction speedup and consequent increases in space-time yields is achieved, while the process window for favorable operation to selectively produce Rufinamide precursor in good yields is widened. The best conditions thus determined are applied to several azide-alkyne cycloadditions to widen the scope of the presented methodology. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  8. Development of a digital card to simulate period transients in research reactors; Desenvolvimento de um cartao digital para simulacao da variacao do periodo em reatores de pesquisa

    Energy Technology Data Exchange (ETDEWEB)

    Masotti, Paulo Henrique Ferraz

    1999-07-01

    This work presents the development of a card to be used in a 'slot' of a micro-computer for evaluation of a nuclear channel used to monitor the start up of nuclear reactors. The results of the bench tests showed good linearity and 2% error deviation in the entire range of operation. Fields tests, performed with the start up channel of IEA-R1 research reactor showed that the card is an excellent device to verify the performance of the channel during steady state, and transient conditions. (author)

  9. Report of the second joint Research Committee for Fusion Reactor and Materials. July 12, 2002, Tokyo, Japan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    Joint research committees in purpose of the discussion on DEMO blanket in view point of the both of reactor technology and materials were held by the Research Committee for Fusion Reactor and Fusion Materials. The joint research committee was held in Tokyo on July 12, 2002. In the committee, the present status of development of solid and liquid breeding blanket, the present status of development of reduced activation structure materials, and IFMIF (International Fusion Materials Irradiation Facility) program were discussed based on the discussions of the development programs of the blanket and materials at the first joint research committee. As a result, it was confirmed that high electric efficiency with 41% would be obtained in the solid breeding blanket system, that neutron radiation data of reduced activation ferritic steel was obtained by HFIR collaboration, and that KEP (key element technology phase) of IFMIF would be finished at the end of 2002 and the data base for the next step, i.e. EVEDA (engineering validation/engineering design activity) was obtained. In addition, the present status of ITER CTA, which was a transient phase for the construction, and the outline of ITER Fast Track, which was an accelerated plan for the performance of the power plants, were reported. This report consists of the summary of the discussion and the viewgraphs which were used at the second joint research committee, and these are very useful for the researchers of the fusion area in Japan. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  11. Impacts of burnup-dependent swelling of metallic fuel on the performance of a compact breed-and-burn fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Heo, Woong; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

    2016-04-15

    The U-Zr or U-TRU-Zr cylindrical metallic fuel slug used in fast reactors is known to swell significantly and to grow during irradiation. In neutronics simulations of metallic-fueled fast reactors, it is assumed that the slug has swollen and contacted cladding, and the bonding sodium has been removed from the fuel region. In this research, a realistic burnup-dependent fuel-swelling simulation was performed using Monte Carlo code McCARD for a single-batch compact sodium-cooled breed-and-burn reactor by considering the fuel-swelling behavior reported from the irradiation test results in EBR-II. The impacts of the realistic burnup-dependent fuel swelling are identified in terms of the reactor neutronics performance, such as core lifetime, conversion ratio, axial power distribution, and local burnup distributions. It was found that axial fuel growth significantly deteriorated the neutron economy of a breed-and-burn reactor and consequently impaired its neutronics performance. The bonding sodium also impaired neutron economy, because it stayed longer in the blanket region until the fuel slug reached 2% burnup.

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

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Mei, E-mail: pm740509@163.com; Sun, Liangwei; Chen, Liang; Sun, Guangai; Chen, Bo; Xie, Chaomei; Xia, Qingzhong; Yan, Guanyun; Tian, Qiang; Huang, Chaoqiang; Pang, Beibei; Zhang, Ying; Wang, Yun; Liu, Yaoguang; Kang, Wu; Gong, Jian

    2016-02-21

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  14. Neutron spectrometry and dosimetry study at two research nuclear reactors using Bonner sphere spectrometer (BSS), rotational spectrometer (ROSPEC) and cylindrical nested neutron spectrometer (NNS).

    Science.gov (United States)

    Atanackovic, J; Matysiak, W; Hakmana Witharana, S S; Aslam, I; Dubeau, J; Waker, A J

    2013-01-01

    Neutron spectrometry and subsequent dosimetry measurements were undertaken at the McMaster Nuclear Reactor (MNR) and AECL Chalk River National Research Universal (NRU) Reactor. The instruments used were a Bonner sphere spectrometer (BSS), a cylindrical nested neutron spectrometer (NNS) and a commercially available rotational proton recoil spectrometer. The purposes of these measurements were to: (1) compare the results obtained by three different neutron measuring instruments and (2) quantify neutron fields of interest. The results showed vastly different neutron spectral shapes for the two different reactors. This is not surprising, considering the type of the reactors and the locations where the measurements were performed. MNR is a heavily shielded light water moderated reactor, while NRU is a heavy water moderated reactor. The measurements at MNR were taken at the base of the reactor pool, where a large amount of water and concrete shielding is present, while measurements at NRU were taken at the top of the reactor (TOR) plate, where there is only heavy water and steel between the reactor core and the measuring instrument. As a result, a large component of the thermal neutron fluence was measured at MNR, while a negligible amount of thermal neutrons was measured at NRU. The neutron ambient dose rates at NRU TOR were measured to be between 0.03 and 0.06 mSv h⁻¹, while at MNR, these values were between 0.07 and 2.8 mSv h⁻¹ inside the beam port and <0.2 mSv h⁻¹ between two operating beam ports. The conservative uncertainty of these values is 15 %. The conservative uncertainty of the measured integral neutron fluence is 5 %. It was also found that BSS over-responded slightly due to a non-calibrated response matrix.

  15. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

  16. Analysis, evaluation, and optimization of kinetic parameters for performance appraisal and design of UASB reactors.

    Science.gov (United States)

    Bhunia, Puspendu; Ghangrekar, M M

    2008-05-01

    Studies have been undertaken to explore the applicability of different kinetic models for the performance appraisal of upflow anaerobic sludge blanket (UASB) reactors treating wastewater in the range of 300-4000 mg COD/l. Three kinetic models namely, Monod, Grau second-order, and Haldane model are considered for the analysis. Both linear and nonlinear regressions have been performed to examine the best-fit among the kinetic models. In this process, five error analysis methods have been used to analyze the data. Apart from optimization of kinetic coefficients with minimization of associated errors, prediction of effluent COD has also been undertaken to verify the applicability of kinetic models. In both the cases, Grau second-order model is found to be the best class of fit for wide range of data sets in UASB reactor.

  17. Studies for a multipurpose research reactor for the CRCN/CNEN-PE

    Energy Technology Data Exchange (ETDEWEB)

    Barroso, Antonio C.O. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Maiorino, Jose R.; Bastos, Jose L.; Silva, Jose E.R. da; Yamaguchi, Mitsuo; Umbehaum, Pedro E.; Ferreira, Carlos R.; Maprelian, E.; Silva, Graciete S. de A.; Yoryiaz, H.; Terremoto, Luiz A.A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: maiorino@net.ipen.br; mitsuo@net.ipen.br; emaprel@net.ipen.br; gsasilva@net.ipen.br; Moreira, Joao M.L. [Centro Tecnologico da Marinha (CTMSP), Sao Paulo, SP (Brazil); Lima, Fernando R. de A.; Lyra, Carlos A.B.O. [Pernambuco Univ., Recife, PE (Brazil). Dept. de Energia Nuclear; Azevedo, Carlos V.G. de; Filho, Jose A.B. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Santos, Rubens S. dos [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil)

    1998-07-01

    This paper presents a conceptual proposal for the design and construction of an irradiation research facility at the CRCN (Regional Center of Nuclear Sciences) site, in cooperation with an international partner. The planned irradiation facility is based on a multipurpose research reactor with an innovative design feature, which is a core with two sub-critical parts coupled by a heavy water tank for enhancing and flatten the thermal fluxes, improving safety, and improving beam applications. (author)

  18. Performance of Aerobic Sequencing Batch reactor (SBR) for Formaldehyde Removal from Synthetic Wastewater

    OpenAIRE

    2013-01-01

    Background and objectives: Formaldehyde is one of the compounds widely used in various industries; hence, its discharge into the effluent is unavoidable. Exposure to formaldehyde has a significant health effects. To prevent these issues, treatment of wastewater containing formaldehyde is necessary. The purpose of this study was to determine the performance of aerobic sequencing batch reactor (SBR) in removing formaldehyde from wastewater. Methods: We used a SBR having a total volume of 6.1...

  19. Analysis of dashpot performance for rotating control drums of a lithium cooled fast reactor concept

    Science.gov (United States)

    Wenzler, C. J.

    1972-01-01

    A dashpot was incorporated in the design of the drive train of the rotating control drum to prevent shock damage to the control drum and drive train at the termination of a scram action. A rotating vane dashpot using reactor coolant lithium as a damping fluid appears to be the best candidate of the various damping devices explored. A performance analysis, results and discussion of vane type dashpots are presented.

  20. Design of Seismic Test Rig for Control Rod Drive Mechanism of Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Jongoh; Kim, Gyeongho; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The reactor assembly is submerged in a reactor pool filled with water and its reactivity is controlled by locations of four control absorber rods(CARs) inside the reactor assembly. Each CAR is driven by a stepping motor installed at the top of the reactor pool and they are connected to each other by a tie rod and an electromagnet. The CARs scram the reactor by de-energizing the electromagnet in the event of a safe shutdown earthquake(SSE). Therefore, the safety function of the control rod drive mechanism(CRDM) which consists of a drive assembly, tie rod and CARs is to drop the CAR into the core within an appropriate time in case of the SSE. As well known, the operability for complex equipment such as the CRDM during an earthquake is very hard to be demonstrated by analysis and should be verified through tests. One of them simulates the reactor assembly and the guide tube of the CAR, and the other one does the pool wall where the drive assembly is installed. In this paper, design of the latter test rig and how the test is performed are presented. Initial design of the seismic test rig and excitation table had its first natural frequency at 16.3Hz and could not represent the environment where the CRDM was installed. Therefore, experimental modal analyses were performed and an FE model for the test rig and table was obtained and tuned based on the experimental results. Using the FE model, the design of the test rig and table was modified in order to have higher natural frequency than the cutoff frequency. The goal was achieved by changing its center of gravity and the stiffness of its sliding bearings.

  1. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  3. Commercial-Scale Performance Predictions for High-Temperature Electrolysis Plants Coupled to Three Advanced Reactor Types

    Energy Technology Data Exchange (ETDEWEB)

    M. G. McKellar; J. E. O' Brien; J. S. Herring

    2007-09-01

    This report presents results of system analyses that have been developed to assess the hydrogen production performance of commercial-scale high-temperature electrolysis (HTE) plants driven by three different advanced reactor – power-cycle combinations: a high-temperature helium cooled reactor coupled to a direct Brayton power cycle, a supercritical CO2-cooled reactor coupled to a direct recompression cycle, and a sodium-cooled fast reactor coupled to a Rankine cycle. The system analyses were performed using UniSim software. The work described in this report represents a refinement of previous analyses in that the process flow diagrams include realistic representations of the three advanced reactors directly coupled to the power cycles and integrated with the high-temperature electrolysis process loops. In addition, this report includes parametric studies in which the performance of each HTE concept is determined over a wide range of operating conditions. Results of the study indicate that overall thermal-to- hydrogen production efficiencies (based on the low heating value of the produced hydrogen) in the 45 - 50% range can be achieved at reasonable production rates with the high-temperature helium cooled reactor concept, 42 - 44% with the supercritical CO2-cooled reactor and about 33 - 34% with the sodium-cooled reactor.

  4. A development of SDLC for MMIS of SMART research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Yong Suk; Park, Jae Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Koo, In Soo; Kim, Jong Myung; Park, Dong Cheol [Control Tech. Research Inst., Ulsan (Korea, Republic of); Kim, Hyeon Soo [Chungnam National University, Taejon (Korea, Republic of)

    2004-07-01

    Software development concept for Man Machine Interface System (MMIS) of SMART has been researched in KAERI since 2000. As a result of it, we developed Software Development Life Cycle (SDLC) based on IEEE Std 1074(1997) and submitted it to Korea Institute of Nuclear Safety (KINS) as a part of Pre-SAR. We verify that the SDLC meets IEEE Std 1074(1997) by making mapping table between them in this paper.

  5. Development and implementation of k{sub 0}-INAA standardization at 10 MW Pakistan research reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Wasim, M.; Arif, M.; Zaidi, J.H.; Anwar, Y. [Chemistry Div., Pakistan Inst. of Nuclear Science and Technology, Islamabad (Pakistan)

    2009-07-01

    The k{sub 0}-method has been developed for 10 MW Pakistan Research Reactor (PARR-1). It involved the full energy peak efficiency calibration of HPGe detector for different counting geometries and characterization of neutron flux at three irradiation channels. Neutron flux was characterized for thermal to epithermal flux ratio, epithermal flux shape factor, modified spectral index. Westcott's g-factor and fast flux. The method was validated by analyzing IAEA-SL1 (lake sediment) and NIST-SRM-1572 (citrus leaves). All calculations were performed in Excel. The results revealed most of the elements with good accuracy. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-03-31

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

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

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2005-01-01

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

  8. Status of research and development on reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    To improve uranium utilization, a design study of the Reduced-Moderation Water Reactor (RMWR) has been carried out intensively since 1998 at the Japan Atomic Energy Research Institute (JAERI). In this reactor, the nuclear fission reaction is designed to be realized mainly by high energy neutrons. To achieve this, the volume of water used to cool the fuel rods is decreased by reducing the gap width between the fuel rods. Conversion ratio greater than 1.0 is expected whether the core i-s cooled by boiling water or pressurized water and whether the core size is small or large. Status of the RMWR design is reviewed and planning of R and D for future deployment of this reactor after 20-20 is presented. To improve economics of this reactor, development of fuel cans for high burnup and low-cost reprocessing technology of mixed oxide spect fuels are highly needed. R and D has been conducted under the cooperation with utilities, industry, research organization and academia. (T. Tanaka)

  9. EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    REID, ROBERT S. [Los Alamos National Laboratory; PEARSON, J. BOSIE [Los Alamos National Laboratory; STEWART, ERIC T. [Los Alamos National Laboratory

    2007-01-16

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  10. Influences of iron and calcium carbonate on wastewater treatment performances of algae based reactors.

    Science.gov (United States)

    Zhao, Zhimiao; Song, Xinshan; Wang, Wei; Xiao, Yanping; Gong, Zhijie; Wang, Yuhui; Zhao, Yufeng; Chen, Yu; Mei, Mengyuan

    2016-09-01

    The influences of iron and calcium carbonate (CaCO3) addition in wastewater treatments reactors performance were investigated. Adding different concentrations of Fe(3+) (5, 10, 30 and 50mmol/m(3)), iron and CaCO3 powder led to changes in algal characteristics and physico-chemical and microbiological properties. According to the investigation results, nutrient removal efficiency in algae based reactors was obviously increased by the addition of 10mmol/m(3) Fe(3+), iron (5mmol/m(3)) and CaCO3 powder (0.2gm(-3)) and the removal efficiencies of BOD5, TN, and TP in Stage 2 were respectively increased by 28%, 8.9%, and 22%. The improvements in physico-chemical performances were verified by microbial community tests (bacteria quantity, activity and community measured in most probable number, extracellular enzymes activity, and Biolog Eco Plates). Microbial variations indicated the coexistence of Fe ions and carbonate-bicarbonate, which triggered the synergistic effect of physico-chemical action and microbial factors in algae based reactors. Copyright © 2016 Elsevier Ltd. All rights reserved.

  11. Performance of a sequencing batch biofilm reactor for the treatment of pre-oxidized sulfamethoxazole solutions.

    Science.gov (United States)

    González, Oscar; Esplugas, Marc; Sans, Carme; Torres, Alicia; Esplugas, Santiago

    2009-05-01

    A combined strategy of a photo-Fenton pretreatment followed by a Sequencing Batch Biofilm Reactor (SBBR) was evaluated for total C and N removal from a synthetic wastewater containing exclusively 200 mg L(-1) of the antibiotic Sulfamethoxazole (SMX). Photo-Fenton reaction was optimized at the minimum reagent doses in order to improve the biocompatibility of effluents with the subsequent biological reactor. Consequently, the pretreatment was performed with two different initial H(2)O(2) concentrations (300 and 400 mg L(-1)) and 10 mg L(-1) of Fe(2+). The pre-treated effluents with the antibiotic intermediates as sole carbon source were used as feed for the biological reactor. The SBBR was operated under aerobic conditions to mineralize the organic carbon, and the Hydraulic Retention Time (HRT) was optimized down to 8h reaching a removal of 75.7% of the initial Total Organic Carbon (TOC). The total denitrification of the NO(3)(-) generated along the chemical-biological treatment was achieved by means of the inclusion of a 24-h anoxic stage in the SBBR strategy. In addition, the Activated Sludge Model No. 1 (ASM1) was successfully used to complete the N balance determining the N fate in the SBBR. The characterization and the good performance of the SBBR allow presenting the assessed combination as an efficient way for the treatment of wastewaters contaminated with biorecalcitrant pharmaceuticals as the SMX.

  12. Treatment of slaughter house wastewater in a sequencing batch reactor: performance evaluation and biodegradation kinetics.

    Science.gov (United States)

    Kundu, Pradyut; Debsarkar, Anupam; Mukherjee, Somnath

    2013-01-01

    Slaughterhouse wastewater contains diluted blood, protein, fat, and suspended solids, as a result the organic and nutrient concentration in this wastewater is vary high and the residues are partially solubilized, leading to a highly contaminating effect in riverbeds and other water bodies if the same is let off untreated. The performance of a laboratory-scale Sequencing Batch Reactor (SBR) has been investigated in aerobic-anoxic sequential mode for simultaneous removal of organic carbon and nitrogen from slaughterhouse wastewater. The reactor was operated under three different variations of aerobic-anoxic sequence, namely, (4+4), (5+3), and (3+5) hr. of total react period with two different sets of influent soluble COD (SCOD) and ammonia nitrogen (NH4(+)-N) level 1000 ± 50 mg/L, and 90 ± 10 mg/L, 1000 ± 50 mg/L and 180 ± 10 mg/L, respectively. It was observed that from 86 to 95% of SCOD removal is accomplished at the end of 8.0 hr of total react period. In case of (4+4) aerobic-anoxic operating cycle, a reasonable degree of nitrification 90.12 and 74.75% corresponding to initial NH4(+)-N value of 96.58 and 176.85 mg/L, respectively, were achieved. The biokinetic coefficients (k, K(s), Y, k(d)) were also determined for performance evaluation of SBR for scaling full-scale reactor in future operation.

  13. Removal performance and microbial communities in a sequencing batch reactor treating hypersaline phenol-laden wastewater.

    Science.gov (United States)

    Jiang, Yu; Wei, Li; Zhang, Huining; Yang, Kai; Wang, Hongyu

    2016-10-01

    Hypersaline phenol-rich wastewater is hard to be treated by traditional biological systems. In this work, a sequencing batch reactor was used to remove phenol from hypersaline wastewater. The removal performance was evaluated in response to the variations of operating parameters and the microbial diversity was investigated by 454 pyrosequencing. The results showed that the bioreactor had high removal efficiency of phenol and was able to keep stable with the increase of initial phenol concentration. DO, pH, and salinity also affected the phenol removal rate. The most abundant bacterial group was phylum Proteobacteria in the two working conditions, and class Gammaproteobacteria as well as Alphaproteobacteria was predominant subgroup. The abundance of bacterial clusters was notably different along with the variation of operation conditions, resulting in changes of phenol degradation rates. The high removal efficiency of phenol suggested that the reactor might be promising in treating phenol-laden industrial wastewater in high-salt condition.

  14. The Effect of Hydrodynamics on Riser Reactor Performance of the FCCU

    Directory of Open Access Journals (Sweden)

    Digieneni Yousuo

    2016-09-01

    Full Text Available The effect of hydrodynamics on the riser reactor performance was studied. The simulation was carried out using COMSOL Multiphysics software. The simulation results showed that reaction rate increased with increase in temperature. The results also showed that an increase in pressure leads to an increases in the velocity of the riser reactor. A maximum on gasoline yield appears when the gas oil inlet temperature is 600K,the catalyst inlet temperature is 1100K and the steam inlet temperature is 400K making gasoline yield between 52% to 55%. A minimum on Coke yield appears when the gas oil inlet temperature is 300K, the catalyst inlet temperature is 800K and the steam inlet temperature is 200K making coke yield between 1% to 2%.

  15. LOCA simulation in the national research universal reactor program: postirradiation examination results for the third materials experiment (MT-3)

    Energy Technology Data Exchange (ETDEWEB)

    Rausch, W.N.

    1984-04-01

    A series of in-reactor experiments were conducted using full-length 32-rod pressurized water reactor (PWR) fuel bundles as part of the Loss-of-Coolant Accident (LOCA) Simulation Program. The third materials experiment (MT-3) was the sixth in the series of thermal-hydraulic and materials deformation/rutpure experiments conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. The main objective of the experiment was to evaluate ballooning and rupture during active two-phase cooling in the temperature range from 1400 to 1500/sup 0/F (1030 to 1090 K). The 12 test rods in the center of the 32-rod bundle were initially pressurized to 550 psi (3.8 MPa) to insure rupture in the correct temperature range. All 12 of the rods ruptured, with an average peak bundle strain of approx. 55%. The UKAEA also funded destructive postirradiation examination (PIE) of several of the ruptured rods from the MT-3 experiment. This report describes the work performed and presents the PIE results. Information obtained during the PIE included cladding thickness measurements metallography, and particle size analysis of the cracked and broken fuel pellets.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  17. Development of a computational database for probabilistic safety assessment of nuclear research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macedo, Vagner S.; Oliveira, Patricia S. Pagetti de; Andrade, Delvonei Alves de, E-mail: vagner.macedo@usp.br, E-mail: patricia@ipen.br, E-mail: delvonei@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The objective of this work is to describe the database being developed at IPEN - CNEN / SP for application in the Probabilistic Safety Assessment of nuclear research reactors. The database can be accessed by means of a computational program installed in the corporate computer network, named IPEN Intranet, and this access will be allowed only to professionals previously registered. Data updating, editing and searching tasks will be controlled by a system administrator according to IPEN Intranet security rules. The logical model and the physical structure of the database can be represented by an Entity Relationship Model, which is based on the operational routines performed by IPEN - CNEN / SP users. The web application designed for the management of the database is named PSADB. It is being developed with MySQL database software and PHP programming language is being used. Data stored in this database are divided into modules that refer to technical specifications, operating history, maintenance history and failure events associated with the main components of the nuclear facilities. (author)

  18. Evaluation of the physical protection system of the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vaz, Antonio C.A.; Conti, Thadeu das N., E-mail: acavaz@ipen.br, E-mail: tnconti@yahoo.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The '09/11' in New York and the accident at the Fukushima power plant are two events that served as worldwide reference to review some aspects of the Physical Protection System (PPS) in nuclear areas. The nuclear research reactor IEA-R1 has followed this new world order and improved the protection systems that are directly related to detection (CCTV, sensors, alarms, etc), delay (turnstile, gates, barriers, etc) and response (communication systems, response force, etc), for operation against malicious act, seeking always to avoid or minimize any possibility of threat, theft and sabotage. These actions were performed to prevent and to mitigate the consequence on the environment, economy and society from damages caused by natural hazard, as well. This study evaluates the PPS of the IEA-R1 regarding the weaknesses, strengths,and impacts of the changes resulting from the system implanted. The analyses were based on methodology developed by security experts from SANDIA National Laboratories in Texas - U.S.A, allowing the evaluation of the system through probabilistic and hypothetical analysis. (author)

  19. Characterization of commercially pure aluminum powder for research reactor fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Downs, V.D. [Babcock and Wilcox Co., Lynchburg, VA (United States); Wiencek, T.C. [Argonne National Lab., IL (United States)

    1992-11-01

    Aluminum powder is used as the matrix material in the production of uranium aluminide, oxide, and silicide dispersion fuel plates for research and test reactors. variability in the characteristics of the aluminum powder, such as moisture content and particle-size distribution, influences blending and compacting of the aluminum/fuel powder. A detailed study was performed to characterize the physical properties of three aluminum powder lots. An angle-of-shear test was devised to characterize the cohesiveness of the aluminum powder. Flow-rate measurements, apparent density determination, subsieve analysis, surface area measurements, and scanning electron microscopy were also used in the study. It was found that because of the various types of commercially available powders, proper specification of powder variables will ensure the receipt of consistent raw materials. Improved control of the initial powder will reduce the variability of fuel-plate production and will improve overall plate reproducibility. It is recommended that a standard specification be written for the aluminum powder and silicide fuel.

  20. Characterization of commercially pure aluminum powder for research reactor fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Downs, V.D. (Babcock and Wilcox Co., Lynchburg, VA (United States)); Wiencek, T.C. (Argonne National Lab., IL (United States))

    1992-01-01

    Aluminum powder is used as the matrix material in the production of uranium aluminide, oxide, and silicide dispersion fuel plates for research and test reactors. variability in the characteristics of the aluminum powder, such as moisture content and particle-size distribution, influences blending and compacting of the aluminum/fuel powder. A detailed study was performed to characterize the physical properties of three aluminum powder lots. An angle-of-shear test was devised to characterize the cohesiveness of the aluminum powder. Flow-rate measurements, apparent density determination, subsieve analysis, surface area measurements, and scanning electron microscopy were also used in the study. It was found that because of the various types of commercially available powders, proper specification of powder variables will ensure the receipt of consistent raw materials. Improved control of the initial powder will reduce the variability of fuel-plate production and will improve overall plate reproducibility. It is recommended that a standard specification be written for the aluminum powder and silicide fuel.