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Sample records for state triga reactor

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

    SciTech Connect

    Douglas Morrell

    2011-03-01

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

  2. Extension of TRIGA reactor capabilities

    SciTech Connect

    Gietzen, A.J.

    1980-07-01

    The first TRIGA reactor went into operation at 10 kW about 22 years ago. Since that time 55 TRIGAs have been put into operation including steady-state powers up to 14,000 kW and pulsing reactors that pulse to 20,000,000 kW. Five more are under construction and a proposal will soon be submitted for a reactor of 25,000 kW. Along with these increases in power levels (and the corresponding fluxes) the experimental facilities have also been expanded. In addition to the installation of new TRIGA reactors with enhanced capabilities many of the older reactors have been modified and upgraded. Also, a number of reactors originally fueled with plate fuel were converted to TRIGA fuel to take advantage of the improved technical and safety characteristics, including the ability for pulsed operation. In order to accommodate increased power and performance the fuel has undergone considerable evolution. Most of the changes have been in the geometry, enrichment and cladding material. However, more recently further development on the UZrH alloy has been carried out to extend the uranium content up to 45% by weight. This increased U content is necessary to allow the use of less than 20% enrichment in the higher powered reactors while maintaining longer core lifetime. The instrumentation and control system has undergone remarkable improvement as the electronics technology has evolved so rapidly in the last two decades. The information display and the circuitry logic has also undergone improvements for enhanced ease of operation and safety. (author)

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

    SciTech Connect

    Marcum, W.R.; Woods, B.G.; Hartman, M.

    2008-07-15

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

  4. Fission product release from TRIGA-LEU reactor fuels

    SciTech Connect

    Baldwin, N.L.; Foushee, F.C.; Greenwood, J.S

    1980-07-01

    Due to present international concerns over nuclear proliferation, TRIGA reactor fuels will utilize only low-enriched uranium (LEU) (enrichment <20%). This requires increased total uranium loading per unit volume of fuel in order to maintain the appropriate fissile loading. Tests were conducted to determine the fractional release of gaseous and metallic fission products from typical uranium-zirconium hydride TRIGA fuels containing up to 45 wt-% uranium. These tests, performed in late 1977 and early 1978, were similar to those conducted earlier on TRIGA fuels with 8.5 wt-% U. Fission gas release measurements were made on prototypic specimens from room temperature to 1100 deg. C in the TRIGA King Furnace Facility. The fuel specimens were irradiated in the TRIGA reactor at a low power level. The fractional releases of the gaseous nuclides of krypton and xenon were measured under steady-state operating conditions. Clean helium was used to sweep the fission gases released during irradiation from the furnace into a standard gas collection trap for gamma counting. The results of these tests on TRIGA-LEU fuel agree well with data from the similar, earlier tests on TRIGA fuel. The correlation used to calculate the release of fission products from 8.5 wt-% U TRIGA fuel applies equally well for U contents up to 45 wt-%. (author)

  5. Results of the MCNP analysis of 20/20 LEU fuel for the Oregon State University TRIGA Reactor

    SciTech Connect

    Dodd, B.; Klein, A.C.; Lewis, B.R.; Merritt, P.A.

    1995-12-31

    The Monte Carlo Neutron/Photon (MCNP) code has been used to perform the neutronics analysis required to support revision of the Oregon State University TRIGA Reactor (OSTR) Safety Analysis Report (SAR). The SAR revision is a necessary part of the preparation of the application for authorization to convert the OSTR core from High Enriched Uranium (HEU) FLIP fuel to a Low Enriched Uranium (LEU) fuel. Before MCNP was applied to LEU-fueled cores, it was first validated by comparing MCNP calculations on FLIP cores to historical, measured values for these cores. The LEU fuel considered was the 20 wt%, 20% enriched (20/20) TRIGA fuel approved by the Nuclear Regulatory Commission (NRC) in NUREG 1282. The results show that the 20/20 fuel is much more reactive than FLIP fuel. A just-critical OSTR FLIP core contains 65 elements, while a just-critical 20/20 core only needs 51 elements. Similarly, the current operational FLIP core consists of 88 elements, whereas a 20/20 core giving the same core excess only requires 65 elements. This presents a significant problem for the OSTR because of potentially significant neutron flux loss in experimental facilities. Further analysis shows that to achieve a full size operational core of about 90 LEU elements the erbium content of the LEU fuel would need to be increased from 0.47wt% to about 0.85 wt%.

  6. Results of the MCNP analysis of 20/20 LEU fuel for the Oregon State University TRIGA reactor

    SciTech Connect

    Dodd, B.; Klein, A.C.; Lewis, B.R.; Merritt, P.A

    1994-07-01

    The Monte Carlo Neutron/Photon (MCNP) code has been used to perform the neutronics analysis required to support revision of the Oregon State University TRIGA Reactor (OSTR) Safety Analysis Report (SAR). The SAR revision is a necessary part of the preparation of the application for authorization to convert the OSTR core from High Enriched Uranium (HEU) FLIP fuel to a Low Enriched Uranium (LEU) fuel. Before MCNP was applied to LEU-fueled cores, it was first validated by comparing MCNP calculations on FLIP cores to historical, measured values for these cores. The LEU fuel considered was the 20 wt%, 20 % enriched (20/20) TRIGA fuel approved by the Nuclear Regulatory Commission (NRC) in NUREG 1282. The results show that the 20/20 fuel is much more reactive than FLIP fuel. A just-critical OSTR FLIP core contains 65 elements, while a just-critical 20/20 core only needs 51 elements. Similarly, the current operational FLIP core consists of 88 elements, whereas a 20/20 core giving the same core excess only requires 65 elements. This presents a significant problem for the OSTR because of potentially significant neutron flux loss in experimental facilities. Further analysis shows that to achieve a full size operational core of about 90 LEU elements the erbium content of the LEU fuel would need to be increased from 0.47 wt% to about 0.85 wt%. (author)

  7. TRIGA research reactor activities around the world

    SciTech Connect

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

    1991-11-01

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

  8. A Computer Code for TRIGA Type Reactors.

    Energy Science and Technology Software Center (ESTSC)

    1992-04-09

    Version 00 TRIGAP was developed for reactor physics calculations of the 250 kW TRIGA reactor. The program can be used for criticality predictions, power peaking predictions, fuel element burn-up calculations and data logging, and in-core fuel management and fuel utilization improvement.

  9. Decommissioning of the Northrop TRIGA reactor

    SciTech Connect

    Cozens, George B.; Woo, Harry; Benveniste, Jack; Candall, Walter E.; Adams-Chalmers, Jeanne

    1986-07-01

    An overview of the administrative and operational aspects of decommissioning and dismantling the Northrop Mark F TRIGA Reactor, including: planning and preparation, personnel requirements, government interfacing, costs, contractor negotiations, fuel shipments, demolition, disposal of low level waste, final survey and disposition of the concrete biological shielding. (author)

  10. Reactor instrumentation renewal of the TRIGA reactor Vienna, Austria

    SciTech Connect

    Boeck, H.; Weiss, H.; Hood, W.E.; Hyde, W.K.

    1992-07-01

    The TRIGA Mark-II reactor at the Atominstitut in Vienna, Austria is replacing its twenty-four year old instrumentation system with a microprocessor based control system supplied by General Atomics. Ageing components, new governmental safety requirements and a need for state of the art instrumentation for training students has spurred the demand for new reactor instrumentation. In Austria a government appointed expert is assigned the responsibility of reviewing the proposed installation and verifying all safety aspects. After a positive review, final assembly and checkout of the instrumentation system may commence. The instrumentation system consists of three basic modules: the control system console, the data acquisition console and the NH-1000 wide range channel. Digital communications greatly reduce interwiring requirements. Hardwired safety channels are independent of computer control, thus, the instrumentation system in no way relies on any computer intervention for safety function. In addition, both the CSC and DAC computers are continuously monitored for proper operation via watchdog circuits which are capable of shutting down the reactor in the event of computer malfunction. Safety channels include two interlocked NMP-1000 multi-range linear channels for steady state mode, an NPP-1000 linear safety channel for pulse mode and a set of three independent fuel temperature monitoring channels. The microprocessor controlled wide range NM- 1000 digital neutron monitor (fission chamber based) functions as a startup/operational channel, and provides all power level related Interlocks. The Atominstitut TRIGA reactor is configured for four modes of operation: manual mode, automatic mode (servo control), pulsing mode and square wave mode. Control of the standard control rods is via stepping motor control rod drives, which offers the operator the choice of which control rods are operated by the servo system in automatic and square wave model. (author)

  11. Thermal-Hydraulic Analysis of the 3-MW TRIGA MARK-II Research Reactor Under Steady-State and Transient Conditions

    SciTech Connect

    Huda, M.Q.; Bhuiyan, S.I.; Chakrobortty, T.K.; Sarker, M.M.; Mondal, M.A.W

    2001-07-15

    Important thermal-hydraulic parameters of the 3-MW TRIGA MARK-II research reactor operating under both steady-state and transient conditions are reported. Neutronic analyses were performed by using the CITATION diffusion code and the MCNP4B2 Monte Carlo code. The output of CITATION and MCNP4B2 were input to the PARET thermal-hydraulic code to study the steady-state and transient thermal-hydraulic behavior of the reactor. To benchmark the PARET model, data were obtained from different measurements performed by thermocouples in the instrumented fuel (IF) rod during the steady-state operation both under forced- and natural-convection mode and compared with the calculation. The mass flow rates needed for input to PARET were taken from the Final Safety Analysis Report for a downward forced coolant flow equivalent to 3500 gal/min. For natural convection cooling of the reactor, the mass flow rate was generated using the NCTRIGA code. Peak fuel temperatures measured by the thermocouples in the IF rods at different power levels of the TRIGA core were compared with the values calculated by PARET. The axial distribution of the temperatures of the fuel centerline, fuel surface, and the cladding surface in the hot channel were calculated for the reactor operating at the full-power level. Fuel surface heat flux and heat transfer coefficients for the hot channel were also calculated for the reactor operating at the full-power level. The investigated results were found to be in good agreement with the experimental and operational values. The testing of the PARET model calculations through benchmarking the available TRIGA experimental and operational data for pulse-mode operations showed that PARET can successfully be used to analyze the transient behavior of the reactor. Major transient parameters, such as peak power and prompt energy released after pulse, full-width at half-maximum of pulse peak, and maximum fuel centerline temperatures for different fuel elements at different

  12. Steady-State Axial Temperature and Flow Velocity in Triga Channel.

    Energy Science and Technology Software Center (ESTSC)

    2007-02-28

    Version 00 TRISTAN-IJS is a computer program for calculating steady-state axial temperature distribution and flow velocity through a vertical coolant channel in low power TRIGA reactor core, cooled by natural circulation. It is designed for steady-state thermohydraulic analysis of TRIGA research reactors operating at a low power level of 1-2 MW.

  13. Monte Carlo modelling of TRIGA research reactor

    NASA Astrophysics Data System (ADS)

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

    2010-10-01

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

  14. Fission-product release from TRIGA-LEU reactor fuels

    SciTech Connect

    Baldwin, N.L.; Foushee, F.C.; Greenwood, J.S.

    1980-11-01

    The release of fission products, both gaseous and volatile metals, from TRIGA fuel is important for the analysis of possible accident conditions related to reactor operation and the design of future TRIGA fuel systems. Because of present national concerns over nuclear proliferation, it has become clear that future reactor fuels will, of necessity, utilize low-enriched uranium (LEU, enrichment <20%). This will require increasing the total uranium loading per unit volume of the higher-loaded TRIGA fuels for the purpose of maintaining the appropriate fissile loading. Because of these new developments, tests were conducted to determine the fractional release of gaseous and metallic fission products from typical uranium-zirconium hydride TRIGA fuels containing 8.5 to 45 wt % uranium.

  15. A 5 MW TRIGA reactor design for radioisotope production

    SciTech Connect

    Veca, Anthony R.; Whittemore, William L.

    1994-07-01

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

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

    SciTech Connect

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

    2006-07-01

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

  17. Neutron fluence and energy reproducibility of a 2-dollar TRIGA reactor Pulse

    SciTech Connect

    Payne, Rosara F.; Drader, Jessica A.; Friese, Judah I.; Greenwood, Lawrence R.; Hines, Corey C.; Metz, Lori A.; Kephart, Jeremy D.; King, Matthew D.; Pierson, Bruce D.; Smith, Jeremy D.; Wall, Donald E.

    2009-10-01

    Washington State University’s 1 MW TRIGA reactor has a long history of utilization for neutron activation analysis (NAA). TRIGA reactors have the ability to pulse, reach supercritical (k>1) for short bursts of time. At this high power and fast time the energy spectrum and neutron fluence are largely uncharacterized. The pulse neutron energy spectrum and fluence were determined by the activation of Cu, Au, Co, Fe, and Ti. These analyses were completed with and without Cd shielding to determine reproducibility between pulses. The applications and implications of the neutron energy and fluence reproducibility to the use of pulsed NAA will be discussed.

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

    NASA Astrophysics Data System (ADS)

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

    2016-05-01

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

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

    PubMed

    Snoj, Luka; Zerovnik, Gašper; Trkov, Andrej

    2012-03-01

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

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

    SciTech Connect

    Moore, M.L.

    1994-01-01

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

  1. 14. U.S. TRIGA users conference. Final program and summary of papers

    SciTech Connect

    1994-07-01

    The following papers were presented at the Conference: Early Development and Use of the TRIGA Reactor; Results of the MCNP Analysis of 20/20 LEU Fuel for the Oregon State University TRIGA Reactor; Upgradeable 2MW TRIGA Reactor Design for the Morocco Nuclear Energy Center McClellan Nuclear Radiation Center TRIGA Reactor: Four Years of Operations.

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

    SciTech Connect

    Ciocaanescu, M.; Ionescu, M.

    1996-08-01

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

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

    SciTech Connect

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

    1982-01-01

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

  4. TRIGA Mark II benchmark experiment; Part I: Steady-state operation

    SciTech Connect

    Mele, I.; Ravnik, M.; Trkov, A. )

    1994-01-01

    The experimental results of startup tests after reconstruction and modification of the TRIGA Mark II reactor in Ljubljana are presented. The experiments were performed with a completely fresh, compact, and uniform core. The operating conditions were well defined and controlled, so that the results can be used as a benchmark test case for TRIGA reactor calculations. Both steady-state and pulse mode operation were tested. In this paper, the following steady-state experiments are treated: critical core and excess reactivity, control rod worths, fuel element reactivity worth distribution, fuel temperature distribution, and fuel temperature reactivity coefficient.

  5. Verifying the Asymmetric Multiple Position Neutron Source (AMPNS) method using the TRIGA reactor

    SciTech Connect

    Kim, Soon-Sam; Leyine, S.H.

    1984-07-01

    A new experimental/analytical method has been developed using the Penn State Breazeale (TRIGA) reactor, to measure the k{sub eff} of a damaged core, e.g., the TMI-2 core, and unfold its k{sub infinity} distribution. This new method, the Asymmetric Multiple Position Neutron Source (AMPNS) method, uses the response of several neutron detectors in fixed positions around the core periphery (and possibly in the core) when a neutron source is placed sequentially in different discrete core positions. Experiments have been performed with the Penn State Breazeale TRIGA Reactor (PSBR) and analyzed with appropriate neutron calculations, using PSU-LEOPARD and EXTERMINATOR-II (EXT-II), to verify the method.

  6. NATCRCTR: One-dimensional thermal-hydraulics analysis code for natural-circulation TRIGA reactors

    SciTech Connect

    Feltus, M.A.; Rubinaccio, G.

    1996-12-31

    The Pennsylvania State University nuclear engineering department is evaluating the upgrade of the Reed College (Portland, Oregon) TRIGA reactor from 250 kW to 1 MW in two areas: thermal-hydraulics and steady-state neutronics analysis. This analysis was initiated as a cooperative effort between Penn State and Reed College as a training project for two International Atomic Energy Agency (IAEA) fellows from Ghana. The two Ghanaian IAEA fellows were assisted by G. Rubinaccio, an undergraduate, who undertook the task of writing the new computer programs for the thermal-hydraulic and physics evaluation as a three-credit special design project course. The Reed College TRIGA, which has a fixed graphite radial reflector, is cooled by natural circulation, without external cross-flow; whereas, the Penn State Breazeale Reactor has significant crossflow into its sides. To model the Reed TRIGA, the NATCRCTR program has been developed from first principles using the following assumptions: 1. The core is surrounded by the fixed reflector structure, which acts as a one-dimensional channel. 2. The core inlet temperature distribution is constant at the core bottom. 3. The axial heat flux distribution is a chopped cosine shape. 4. The heat transfer in the fuel is primarily in the radial directions. 5. A small gap between the fuel and cladding exists. The NATCRCTR code is used to find the peak centerline fuel, gap, and cladding surface temperatures, based on assumed flux and engineering peaking factors.

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

    SciTech Connect

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-22

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

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

    NASA Astrophysics Data System (ADS)

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-01

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

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

    NASA Astrophysics Data System (ADS)

    Ramli, Nurhayati; Maskin, Mazleha; Tom, Phongsakorn Prak; Husain, Nurfazila; Farid, Mohd Fairus Abd; Ramli, Shaharum; Adnan, Amirul Syazwan; Abidin, Nurul Husna Zainal

    2016-01-01

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

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

    SciTech Connect

    Ionita, I. Grabcev, B.; Todireanu, S.; Constantin, F.; Shvetsov, V.; Anghel, E.; Popescu, G.; Mincu, M.; Datcu, A.

    2006-12-15

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

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

    NASA Astrophysics Data System (ADS)

    Erradi, L.; Essadki, H.

    2001-06-01

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

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

    SciTech Connect

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

    1991-01-01

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

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

    SciTech Connect

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

    1991-12-31

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

  14. 78 FR 26811 - Dow Chemical Company, Dow TRIGA Research Reactor; License Renewal for the Dow Chemical TRIGA...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-08

    ...) published a notice in the Federal Register on July 20, 2012 (77 FR 42771), ``License Renewal for the Dow...: I. Correction In the Federal Register (FR) of July 20, 2012, in FR Doc. 2012- 17733, on page 42772... COMMISSION Dow Chemical Company, Dow TRIGA Research Reactor; License Renewal for the Dow Chemical...

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

    Energy Science and Technology Software Center (ESTSC)

    2001-08-29

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

  16. Deployment of a three-dimensional array of Micro-Pocket Fission Detector triads (MPFD3) for real-time, in-core neutron flux measurements in the Kansas State University TRIGA Mark-II Nuclear Reactor

    NASA Astrophysics Data System (ADS)

    Ohmes, Martin Francis

    A Micro-Pocket Fission Detector (MPFD) is a miniaturized type of fission chamber developed for use inside a nuclear reactor. Their unique design allows them to be located between or even inside fuel pins while being built from materials which give them an operational lifetime comparable to or exceeding the life of the fuel. While other types of neutron detectors have been made for use inside a nuclear reactor, the MPFD is the first neutron detector which can survive sustained use inside a nuclear reactor while providing a real-time measurement of the neutron flux. This dissertation covers the deployment of MPFDs as a large three-dimensional array inside the Kansas State University TRIGA Mark-II Nuclear Reactor for real-time neutron flux measurements. This entails advancements in the design, construction, and packaging of the Micro-Pocket Fission Detector Triads with incorporated Thermocouple, or MPFD3-T. Specialized electronics and software also had to be designed and built in order to make a functional system capable of collecting real-time data from up to 60 MPFD3-Ts, or 180 individual MPFDs and 60 thermocouples. Design of the electronics required the development of detailed simulations and analysis for determining the theoretical response of the detectors and determination of their size. The results of this research shows that MPFDs can operate for extended times inside a nuclear reactor and can be utilized toward the use as distributed neutron detector arrays for advanced reactor control systems and power mapping. These functions are critical for continued gains in efficiency of nuclear power reactors while also improving safety through relatively inexpensive redundancy.

  17. Technical Specifications for the Neutron Radiography Facility (TRIGA Mark 1 Reactor). Revision 6

    SciTech Connect

    Tomlinson, R.L.; Perfect, J.F.

    1988-04-01

    These Technical Specifications state the limits under which the Neutron Radiography Facility, with its associated TRIGA Mark I Reactor, is operated by the Westinghouse Hanford Company for the US Department of Energy. These specifications cover operation of the Facility for the purpose of examination of specimens (including contained fissile material) by neutron radiography, for the irradiation of specimens in the pneumatic transfer system and approved in-core or in-pool irradiation facilities and operator training. The Final Safety Analysis Report (TC-344) and its supplements, and these Technical Specifications are the basic safety documents of the Neutron Radiography Facility.

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

    SciTech Connect

    Sholtis, Joseph A. Jr.

    1982-07-01

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

  19. Environmental Assessment: Relocation and storage of TRIGA{reg_sign} reactor fuel, Hanford Site, Richland, Washington

    SciTech Connect

    1995-08-01

    In order to allow the shutdown of the Hanford 308 Building in the 300 Area, it is proposed to relocate fuel assemblies (101 irradiated, three unirradiated) from the Mark I TRIGA Reactor storage pool. The irradiated fuel assemblies would be stored in casks in the Interim Storage Area in the Hanford 400 Area; the three unirradiated ones would be transferred to another TRIGA reactor. The relocation is not expected to change the offsite exposure from all Hanford Site 300 and 400 Area operations.

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

    SciTech Connect

    Heidel, C.C.; Richards, W.J.

    1994-07-01

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

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

    NASA Astrophysics Data System (ADS)

    Alkaabi, Ahmed K.

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

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

    SciTech Connect

    Forsbacka, Matt

    1990-07-01

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

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

    NASA Astrophysics Data System (ADS)

    Shugart, Nicolas

    The United States Geological Survey TRIGA Reactor (GSTR) is a 1 MW reactor located in Lakewood, Colorado. In support of the GSTR's relicensing efforts, this project developed and validated a Monte Carlo N-Particle Version 5 (MCNP5) model of the GSTR reactor. The model provided estimates of the excess reactivity, power distribution and the fuel temperature, water temperature, void, and power reactivity coefficients for the current and limiting core. The MCNP5 model predicts a limiting core excess reactivity of 6.48 with a peak rod power of 22.2 kW. The fuel and void reactivity coefficients for the limiting core are strongly negative, and the core water reactivity coefficient is slightly positive, consistent with other TRIGA analyses. The average fuel temperature reactivity coefficient of the full power limiting core is -0.0135 /K while the average core void coefficient is -0.069 /K from 0-20 % void. The core water temperature reactivity coefficient is +0.012 /K. Following the neutronics analysis, the project developed RELAP5 and PARET-ANL models of the GSTR hot-rod fuel channel under steady state and transient conditions. The GSTR limiting core, determined as part of this analysis, provides a worst case operating scenario for the reactor. During steady state operations, the hot rod of the limiting core has a peak fuel temperature of 829 K and a minimum departure from nucleate boiling ratio of 2.16. After a $3.00 pulse reactivity insertion the fuel reaches a peak temperature is 1070 K. Examining the model results several seconds after a pulse reveals flow instabilities that result from weaknesses in the current two-channel model.

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

    SciTech Connect

    Moore, M.L.; Elsasser, S.

    1986-05-01

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

  5. 78 FR 5840 - Notice of License Termination for University of Illinois Advanced TRIGA Reactor, License No. R-115

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-01-28

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

  6. Analysis of decommissioning costs for the AFRRI TRIGA reactor facility. Technical report

    SciTech Connect

    Forsbacka, M.; Moore, M.

    1989-12-01

    This report provides a cost analysis for decommissioning the Armed Forces Radiobiology Research Institute (AFRRI) TRIGA reactor facility. AFRRI is not suggesting that the AFRRI TRIGA reactor facility be decommissioned. This report was prepared in compliance with paragraph 50.33 of Title 10, Code of Federal Regulations, which requires that funding for the decommissioning of reactor facilities be available when licensed activities cease. The planned method of decommissioning is complete decontamination (DECON) of the AFRRI TRIGA reactor site to allow for restoration of the site to full public access. The cost of DECON in 1990 dollars is estimated to be $3,200,000. The anticipated ancillary costs of facility site demobilization and spent fuel shipment will be an additional $600,000. Thus, the total cost of terminating reactor operations at AFRRI will be about $3,800,000. The primary basis for developing this cost estimate was a study of the decommissioning costs of similar reactor facility performed by Battelle Pacific Northwest Laboratory, as provided in U.S. Nuclear Regulatory Commission publication NUREG/CR-1756. The data in this study were adapted to reflect the decommissioning requirements of the AFRRI TRIGA reactor facility.

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

    PubMed

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

    2015-10-01

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

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

    NASA Astrophysics Data System (ADS)

    Fiantini, Rosalina; Umar, Efrizon

    2010-06-01

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

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

    SciTech Connect

    Fiantini, Rosalina; Umar, Efrizon

    2010-06-22

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

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

    NASA Astrophysics Data System (ADS)

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

    2014-02-01

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

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

    SciTech Connect

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

    2014-02-12

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

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

    SciTech Connect

    Li, S.-H.; Shiau, L.-C.

    1990-07-01

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

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

    SciTech Connect

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

    2014-02-12

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

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

    NASA Astrophysics Data System (ADS)

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

    2014-02-01

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

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

    SciTech Connect

    Whittemore, W.L. )

    1993-01-01

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

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

    SciTech Connect

    Snoj, L.; Stancar, Z.; Radulovic, V.; Podvratnik, M.; Zerovnik, G.; Trkov, A.; Barbot, L.; Domergue, C.; Destouches, C.

    2012-07-01

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

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

    PubMed

    Henry, R; Tiselj, I; Snoj, L

    2015-03-01

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

  18. Production of {sup 99}Mo using LEU and molybdenum targets in a 1 MW Triga reactor

    SciTech Connect

    Mo, S.C.

    1993-12-31

    The production of {sup 99}Mo using Low Enriched Uranium (LEU) and natural molybdenum targets in a 1 MW Triga reactor is investigated. The successive linear programming technique is applied to minimize the target loadings for different yield constraints. The irradiation time is related to the kinetics of the growth and decay of {sup 99}Mo. The feasibility of a neutron generated based {sup 99}Mo production system is discussed.

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

    PubMed

    Burgio, N; Rosa, R

    2004-10-01

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

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

    PubMed

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

    2015-09-01

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

  1. New fuel management plan for the Penn State TRIGA

    SciTech Connect

    Hughes, D.; Boyle, P.; Levine, S.H.

    1996-12-31

    The Pennsylvania State University (PSU) Breazeale TRIGA has utilized 12 wt% U fuel in the core since July 1992, when six 12 wt% U fuel elements were loaded to replace the depleted 8.5 wt% U fuel in the centermost ring, the B ring. This reload increased the cold k{sub eff} from 1.03 to 1.05, the cold k{sub eff} of 1.03 being the minimum k{sub eff} that will permit 1-MW operation for a sustained period. In the next fuel reload, this 12 wt% U fuel is to be moved outward to the adjacent ring, the C ring, and six fresh 12 wt% U fuel elements are to be added to the B ring. It was determined that using the 12 wt% U in place of 8.5 wt% U fuel reduced fuel costs by a factor of 6, and continuing this use of six 12 wt% U fuel elements for each reload maintained the lower fuel costs. This reloading technique worked successfully, requiring only 26 additional 12 wt% U elements to be loaded into the core during the last 23 yr. Recently, however, a new instrumented 12 wt% U fuel element read much higher temperatures than all previous similar fuel elements. Its measured fuel temperature at 1 MW is 585{degrees}C. As a result, the PSU TRIGA now operates at or below 60% full power to prevent this element from reaching fuel temperatures well above 500{degrees}C. The purpose of this paper is to describe a new fuel management strategy developed to use 12 wt% U fuel, which permits 1-MW operation and limits the maximum fuel temperature to {approx}500{degrees}C.

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

    NASA Astrophysics Data System (ADS)

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

    2014-04-01

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

  3. Startup experience at the University of Texas TRIGA Mark II Reactor

    SciTech Connect

    Bauer, Thomas L.; Wehring, Bernard W.

    1992-07-01

    After eight years of singular effort, the UT-TRIGA Mark II research reactor was licensed and is fully operational. This reactor is the focus of a new reactor laboratory facility which is located at the Balcones Research Center, a north Austin campus of The University of Texas at Austin. The UT-TRIGA reactor is licensed for 1.1 MW steady power operation and 3 dollar pulsing. A startup program was implemented upon receipt of the facility license on January 17, 1992. Several facility features are unique to this startup. Among these were the use of fuel with various burnup and a digital control system. The reactor laboratory staff with assistance from a General Atomics instrumentation engineer performed all phases of the startup program. Core loading began in February 1992 with final testing completed in May 1992. Several unusual problems were encountered during this time. Experiment authorizations have been written to resume Neutron Activation Analysis programs and isotope production. Several neutron beam tube experiments are in the design and test phase. (author)

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

    PubMed

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

    2015-02-01

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

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

    SciTech Connect

    Rivero-Gutierrez, T.

    2006-07-01

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

  6. Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania's 14-MW TRIGA reactor

    SciTech Connect

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

    1992-01-01

    The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, each containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations.

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

    PubMed

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

    2015-05-01

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

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

    PubMed

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

    2014-01-01

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

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

    SciTech Connect

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

    2006-07-01

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

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

    SciTech Connect

    Denton, Michael M.; Lim, Tek H.

    1986-07-01

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

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

    SciTech Connect

    Aguilar, H.F.; Mazon, R.R.

    1994-07-01

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

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

    NASA Astrophysics Data System (ADS)

    Viitanen, Tuomas; Leppänen, Jaakko

    2016-02-01

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

  13. NRF TRIGA packaging

    SciTech Connect

    Clements, M.D.

    1995-11-01

    Training Reactor Isotopes, General Atomics (TRIGA{reg_sign}) Reactors are in use at four US Department of Energy (DOE) complex facilities and at least 23 university, commercial, or government facilities. The development of the Neutron Radiography Facility (NRF) TRIGA packaging system began in October 1993. The Hanford Site NRF is being shut down and requires an operationally user-friendly transportation and storage packaging system for removal of the TRIGA fuel elements. The NRF TRIGA packaging system is designed to remotely remove the fuel from the reactor and transport the fuel to interim storage (up to 50 years) on the Hanford Site. The packaging system consists of a cask and an overpack. The overpack is used only for transport and is not necessary for storage. Based upon the cask`s small size and light weight, small TRIGA reactors will find it versatile for numerous refueling and fuel storage needs. The NRF TRIGA packaging design also provides the basis for developing a certifiable and economical packaging system for other TRIGA reactor facilities. The small size of the NRF TRIGA cask also accommodates placing the cask into a larger certified packaging for offsite transport. The Westinghouse Hanford Company NRF TRIGA packaging, as described herein can serve other DOE sites for their onsite use, and the design can be adapted to serve university reactor facilities, handling a variety of fuel payloads.

  14. 76 FR 69296 - University of Utah, University of Utah TRIGA Nuclear Reactor, Notice of Issuance of Renewed...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-11-08

    ... published in the Federal Register on July 21, 2011 (76 FR 43733-43737). The NRC received no request for a..., 2011 (76 FR 60091-60094), and concluded that renewal of the facility operating license will not have a... COMMISSION University of Utah, University of Utah TRIGA Nuclear Reactor, Notice of Issuance of...

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

    NASA Astrophysics Data System (ADS)

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

    2016-01-01

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  16. Fundamental approach to TRIGA steady-state thermal-hydraulic CHF analysis.

    SciTech Connect

    Feldman, E.; Nuclear Engineering Division

    2008-03-30

    Methods are investigated for predicting the power at which critical heat flux (CHF) occurs in TRIGA reactors that rely on natural convection for primary flow. For a representative TRIGA reactor, two sets of functions are created. For the first set, the General Atomics STAT code and the more widely-used RELAP5-3D code are each employed to obtain reactor flow rate as a function of power. For the second set, the Bernath correlation, the 2006 Groeneveld table, the Hall and Mudawar outlet correlation, and each of the four PG-CHF correlations for rod bundles are used to predict the power at which CHF occurs as a function of channel flow rate. The two sets of functions are combined to yield predictions of the power at which CHF occurs in the reactor. A combination of the RELAP5-3D code and the 2006 Groeneveld table predicts 67% more CHF power than does a combination of the STAT code and the Bernath correlation. Replacing the 2006 Groeneveld table with the Bernath CHF correlation (while using the RELAP5-3D code flow solution) causes the increase to be 23% instead of 67%. Additional RELAP5-3D flow-versus-power solutions obtained from Reference 1 and presented in Appendix B for four specific TRIGA reactors further demonstrates that the Bernath correlation predicts CHF to occur at considerably lower power levels than does the 2006 Groeneveld table. Because of the lack of measured CHF data in the region of interest to TRIGA reactors, none of the CHF correlations considered can be assumed to provide the definitive CHF power. It is recommended, however, to compare the power levels of the potential limiting rods with the power levels at which the Bernath and 2006 Groeneveld CHF correlations predict CHF to occur.

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

    NASA Astrophysics Data System (ADS)

    Koehl, Michael A.

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

  18. Reduced enrichment neutronics evaluation for Texas A and M's TRIGA reactor

    SciTech Connect

    Rajalakshmi, M.J.; Reuscher, J.A. )

    1990-06-01

    The Texas A and M Nuclear Science Center reactor (NSCR) designed by General Atomics (GA) uses a fuel-life improvement program TRIGA fuel element. It is composed of 8.5 wt% uranium in U-ZrH{sub 1.6}-Er fuel with a {sup 235}U enrichment of 70 at.% and 1.5 wt% of erbium. The US Nuclear Regulatory Commission requires that the enrichment not exceed 20 at.% for the next loading, provided the fuel is available. To meet this requirement GA has developed shrouded four-rod clusters using low-enriched U-ZrH{sub 1.6}-Er fuel for TRIGA cores operating at powers up to 2 MW. This fuel contains 20 wt% of uranium with an enrichment of 20 at.% and 0.5 wt% of erbium and a homogeneous mixture of hydrogen moderator. Thermal-hydraulic calculations show the feasibility of operating the NSCR at 2 MW. The objective of this study is to assess the ability of the NSCR to operate at 2 MW using the reduced-enrichment fuel. This study also covers a three-dimensional neutronics analysis of the NSCR core using the new fuel. Results obtained are compared with results obtained with another candidate fuel, BeO-UO{sub 2}-Er.

  19. Electron versus proton accelerator driven sub-critical system performance using TRIGA reactors at power

    SciTech Connect

    Carta, M.; Burgio, N.; D'Angelo, A.; Santagata, A.; Petrovich, C.; Schikorr, M.; Beller, D.; Felice, L. S.; Imel, G.; Salvatores, M.

    2006-07-01

    This paper provides a comparison of the performance of an electron accelerator-driven experiment, under discussion within the Reactor Accelerator Coupling Experiments (RACE) Project, being conducted within the U.S. Dept. of Energy's Advanced Fuel Cycle Initiative (AFCI), and of the proton-driven experiment TRADE (TRIGA Accelerator Driven Experiment) originally planned at ENEA-Casaccia in Italy. Both experiments foresee the coupling to sub-critical TRIGA core configurations, and are aimed to investigate the relevant kinetic and dynamic accelerator-driven systems (ADS) core behavior characteristics in the presence of thermal reactivity feedback effects. TRADE was based on the coupling of an upgraded proton cyclotron, producing neutrons via spallation reactions on a tantalum (Ta) target, with the core driven at a maximum power around 200 kW. RACE is based on the coupling of an Electron Linac accelerator, producing neutrons via photoneutron reactions on a tungsten-copper (W-Cu) or uranium (U) target, with the core driven at a maximum power around 50 kW. The paper is focused on analysis of expected dynamic power response of the RACE core following reactivity and/or source transients. TRADE and RACE target-core power coupling coefficients are compared and discussed. (authors)

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

    NASA Astrophysics Data System (ADS)

    Žagar, Tomaž; Božič, Matjaž; Ravnik, Matjaž

    2004-12-01

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

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

    NASA Technical Reports Server (NTRS)

    Miley, G. H.

    1979-01-01

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

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

    SciTech Connect

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

    2014-09-03

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

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

    NASA Astrophysics Data System (ADS)

    O'Kelly, David Sean

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

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

    PubMed

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

    2012-10-01

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

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

    NASA Astrophysics Data System (ADS)

    Bărbos, D.; Păunoiu, C.; Roth, C.

    2010-10-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2014-09-01

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

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

    NASA Astrophysics Data System (ADS)

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

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

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

    NASA Astrophysics Data System (ADS)

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

    2006-08-01

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

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

    SciTech Connect

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

    2015-07-23

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-07-01

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

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

    NASA Astrophysics Data System (ADS)

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

    1993-10-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2013-06-01

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

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

    SciTech Connect

    Ziegner, Markus; Schmitz, Tobias; Hampel, Gabriele; Khan, Rustam; Blaickner, Matthias; Palmans, Hugo; Sharpe, Peter; Böck, Helmuth

    2014-11-01

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

  14. Design, construction, and demonstration of a neutron beamline and a neutron imaging facility at a Mark-I TRIGA reactor

    NASA Astrophysics Data System (ADS)

    Craft, Aaron E.

    The fleet of research and training reactors is aging, and no new research reactors are planned in the United States. Thus, there is a need to expand the capabilities of existing reactors to meet users' needs. While many research reactors have beam port facilities, the original design of the United States Geological Survey TRIGA Reactor (GSTR) did not include beam ports. The MInes NEutron Radiography (MINER) facility developed by this thesis and installed at the GSTR provides new capabilities for both researchers and students at the Colorado School of Mines. The facility consists of a number of components, including a neutron beamline and beamstop, an optical table, an experimental enclosure and associated interlocks, a computer control system, a multi-channel plate imaging detector, and the associated electronics. The neutron beam source location, determined through Monte Carlo modeling, provides the best mixture of high neutron flux, high thermal neutron content, and low gamma radiation content. A Monte Carlo n-Particle (MCNP) model of the neutron beam provides researchers with a tool for designing experiments before placing objects in the neutron beam. Experimental multi-foil activation results, compared to calculated multi-foil activation results, verify the model. The MCNP model predicts a neutron beamline flux of 2.2*106 +/- 6.4*105 n/cm2-s based on a source particle rate determined from the foil activation experiments when the reactor is operating at a power of 950 kWt with the beam shutter fully open. The average cadmium ratio of the beamline is 7.4, and the L/D of the neutron beam is approximately 200+/-10. Radiographs of a sensitivity indicator taken using both the digital detector and the transfer foil method provide one demonstration of the radiographic capabilities of the new facility. Calibration fuel pins manufactured using copper and stainless steel surrogate fuel pellets provide additional specimens for demonstration of the new facility and offer a

  15. Fundamental approach to TRIGA steady-state thermal-hydraulic CHF analysis

    SciTech Connect

    Feldman, E.E.

    2008-07-15

    Methods are investigated for predicting the power at which critical heat flux (CHF) occurs in TRIGA reactors that rely on natural convection for primary flow. For a representative TRIGA reactor, two sets of functions are created. For the first set, the General Atomics STAT code and the more widely-used RELAP5-3D code are each employed to obtain reactor flow rate as a function of power. For the second set, the Bernath correlation, the 2006 Groeneveld table, the Hall and Mudawar outlet correlation, and each of the four PG-CHF correlations for rod bundles are used to predict the power at which CHF occurs as a function of channel flow rate. The two sets of functions are combined to yield predictions of the power at which CHF occurs in the reactor. A combination of the RELAP5-3D code and the 2006 Groeneveld table predicts 67% more CHF power than does a combination of the STAT code and the Bernath correlation. (author)

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

    PubMed

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

    2016-10-01

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

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

    PubMed

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

    2011-11-01

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

  18. Conversion of TRIGA research reactors from high-enriched- to low-enriched-uranium fuels: owner/operator view

    SciTech Connect

    Feltz, D.E.

    1986-01-01

    In June 1985, the US Nuclear Regulatory Commission (NRC) commissioners issued a four-point directive to the NRC staff concerning the conversion of research reactors from the use of high-enriched-uranium (HEU) to low-enriched-uranium (LEU) fuels. As a result of this directive, the earlier concerns of the research reactor community that were presented to the NRC during the comment period of the 1984 proposed rule on HEU-LEU conversion must be dealt with now. This paper discusses the items of most concern to HEU TRIGA owner/operators for conversion to LEU fuel.

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

    NASA Astrophysics Data System (ADS)

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

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

  20. Critical heat flux in natural convection cooled TRIGA reactors with hexagonal bundle

    SciTech Connect

    Yang, J.; Avery, M.; De Angelis, M.; Anderson, M.; Corradini, M.; Feldman, E. E.; Dunn, F. E.; Matos, J. E.

    2012-07-01

    A three-rod bundle Critical Heat Flux (CHF) study at low flow, low pressure, and natural convection condition has been conducted, simulating TRIGA reactors with the hexagonally configured core. The test section is a custom-made trefoil shape tube with three identical fuel pin heater rods located symmetrically inside. The full scale fuel rod is electrically heated with a chopped-cosine axial power profile. CHF experiments were carried out with the following conditions: inlet water subcooling from 30 K to 95 K; pressure from 110 kPa to 230 kPa; mass flux up to 150 kg/m{sup 2}s. About 50 CHF data points were collected and compared with a few existing CHF correlations whose application ranges are close to the testing conditions. Some tests were performed with the forced convection to identify the potential difference between the CHF under the natural convection and forced convection. The relevance of the CHF to test parameters is investigated. (authors)

  1. Production of 37Ar in The University of Texas TRIGA reactor facility

    SciTech Connect

    Egnatuk, Christine M.; Lowrey, Justin; Biegalski, S.; Bowyer, Ted W.; Haas, Derek A.; Orrell, John L.; Woods, Vincent T.; Keillor, Martin E.

    2011-06-19

    The detection of {sup 37}Ar is important for on-site inspections for the Comprehensive Nuclear-Test-Ban Treaty monitoring. In an underground nuclear explosion this radionuclide is produced by {sup 40}Ca(n,{alpha}){sup 37}Ar reaction in surrounding soil and rock. With a half-life of 35 days, {sup 37}Ar provides a signal useful for confirming the location of an underground nuclear event. An ultra-low-background proportional counter developed by Pacific Northwest National Laboratory is used to detect {sup 37}Ar, which decays via electron capture. The irradiation of Ar gas at natural enrichment in the 3L facility within the Mark II TRIGA reactor facility at The University of Texas at Austin provides a source of {sup 37}Ar for the calibration of the detector. The {sup 41}Ar activity is measured by the gamma activity using an HPGe detector after the sample is removed from the core. Using the {sup 41}Ar/{sup 37}Ar production ratio and the {sup 41}Ar activity, the amount of {sup 37}Ar created is calculated. The {sup 41}Ar decays quickly (half-life of 109.34 minutes) leaving a radioactive sample of high purity {sup 37}Ar and only trace levels of {sup 39}Ar.

  2. Analysis of higher than normal fuel temperatures in the hexagonal geometry TRIGA reactor

    SciTech Connect

    Hughes, D.; Boyle, P.; Levine, S.H.

    1996-12-31

    The 1-MW Pennsylvania State University TRIGA has hexagonal geometry and a water-filled central thimble. It was operated with all 8.5 wt% U fuel from December 1965 until July 1972, when 12 wt% U fuel elements replaced the 8.5 wt% U fuel in the centermost ring, the B ring. Although the power density of the 12 wt% U fuel was {approximately}35% greater than the corresponding 8.5 wt% U fuel, its maximum steady-state fuel temperature was always below 500{degrees}C when operating at 1 MW. Since that time, the core has been successfully loaded by placing six 12 wt% U fuel elements in the B ring. The used 12 wt% U fuel moved outward. Recently, however, a new instrumented 12 wt% U fuel element initially read a much higher fuel temperature than all previous similar fuel elements. The purpose of this paper is to present the calculations and experiments performed to correlate calculations with experimental data and to determine the cause of the higher fuel temperature for this element.

  3. Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania`s 14-MW TRIGA reactor

    SciTech Connect

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

    1992-12-01

    The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, each containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations.

  4. Conceptual Design of a Clinical BNCT Beam in an Adjacent Dry Cell of the Jozef Stefan Institute TRIGA Reactor

    SciTech Connect

    Maucec, Marko

    2000-11-15

    The MCNP4B Monte Carlo transport code is used in a feasibility study of the epithermal neutron boron neutron capture therapy facility in the thermalizing column of the 250-kW TRIGA Mark II reactor at the Jozef Stefan Institute (JSI). To boost the epithermal neutron flux at the reference irradiation point, the efficiency of a fission plate with almost 1.5 kg of 20% enriched uranium and 2.3 kW of thermal power is investigated. With the same purpose in mind, the TRIGA reactor core setup is optimized, and standard fresh fuel elements are concentrated partly in the outermost ring of the core. Further, a detailed parametric study of the materials and dimensions for all the relevant parts of the irradiation facility is carried out. Some of the standard epithermal neutron filter/moderator materials, as well as 'pressed-only' low-density Al{sub 2}O{sub 3} and AlF{sub 3}, are considered. The proposed version of the BNCT facility, with PbF{sub 2} as the epithermal neutron filter/moderator, provides an epithermal neutron flux of {approx}1.1 x 10{sup 9} n/cm{sup 2}.s, thus enabling patient irradiation times of <60 min. With reasonably low fast neutron and photon contamination ([overdot]D{sub nfast}/{phi}{sub epi} < 5 x 10{sup -13} Gy.cm{sup 2}/n and [overdot]D{sub {gamma}} /{phi}{sub epi} < 3 x 10{sup -13} Gy.cm{sup 2}/n), the in-air performances of the proposed beam are comparable to all existing epithermal BNCT facilities. The design presents an equally efficient alternative to the BNCT beams in TRIGA reactor thermal columns that are more commonly applied. The cavity of the dry cell, a former JSI TRIGA reactor spent-fuel storage facility, adjacent to the thermalizing column, could rather easily be rearranged into a suitable patient treatment room, which would substantially decrease the overall developmental costs.

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

    SciTech Connect

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

    2014-01-01

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

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

    PubMed

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

    2016-03-01

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

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

    NASA Astrophysics Data System (ADS)

    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

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

    NASA Astrophysics Data System (ADS)

    Esen, Ayse Nur; Haciyakupoglu, Sevilay

    2016-02-01

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

  9. OREGON STATE UNIVERSITY (OSU) TRAINING RESEARCH ISOTOPE GENERAL ATOMICS (TRIGA) OVERPACK CLOSURE WELDING PROCESS PARAMETER DEVELOPMENT & QUALIFICATION

    SciTech Connect

    CANNELL, G.R.

    2006-09-11

    Spent Nuclear Fuel (SNF) from the Oregon State University (OSU) TRIGA{reg_sign} Reactor is currently being stored in thirteen 55-gallon drums at the Hanford Site's low-level burial grounds. This fuel is soon to be retrieved from buried storage and packaged into new containers (overpacks) for interim storage at the Hanford Interim Storage Area (ISA). One of the key activities associated with this effort is final closure of the overpack by welding. The OSU fuel is placed into an overpack, a head inserted into the overpack top, and welded closed. Weld quality, for typical welded fabrication, is established through post-weld testing and nondestructive examination (NDE); however, in this case, once the SNF is placed into the overpack, routine testing and NDE are not feasible. An alternate approach is to develop and qualify the welding process/parameters, demonstrate beforehand that they produce the desired weld quality, and then verify parameter compliance during production welding. Fluor engineers have developed a Gas Tungsten Arc Welding (GTAW) technique and parameters, demonstrating that weld quality requirements for closure of packaged SNF overpacks are met, using this alternate approach. The following reviews the activities performed for this development and qualification effort.

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

    PubMed

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

    2009-07-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-01-01

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

  12. Calculation of the Activity Inventory for the TRIGA Reactor at the Medical University of Hannover (MHH) in Preparation for Dismantling the Facility

    SciTech Connect

    Hampel, G.; Scheller, F.; Bernnat, W.; Pfister, G.; Klaux, U.; Gerhards, E.

    2002-02-25

    It is planned to dismantle the TRIGA reactor facility at the Medical University of Hannover (MHH). Radioactive waste resulting from this dismantling will be disposed of externally, any remaining materials as well as the building structures will then be measured to ensure there is no residual activity. In preparation for this and to plan the techniques which will be used to dismantle the reactor, calculations were made in order to determine the amount of activity and the dose rates for the reactor tank and its inside components as well as for the biological shield and its radial beam tube.

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

    SciTech Connect

    Matos, J.E.

    1994-12-01

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

  14. Measurements of miniature ionization chamber currents in the JSI TRIGA Mark II reactor demonstrate the importance of the delayed contribution to the photon field in nuclear reactors

    NASA Astrophysics Data System (ADS)

    Radulović, Vladimir; Fourmentel, Damien; Barbot, Loïc; Villard, Jean-François; Kaiba, Tanja; Gašper, Žerovnik; Snoj, Luka

    2015-12-01

    The characterization of experimental locations of a research nuclear reactor implies the determination of neutron and photon flux levels within, with the best achievable accuracy. In nuclear reactors, photon fluxes are commonly calculated by Monte Carlo simulations but rarely measured on-line. In this context, experiments were conducted with a miniature gas ionization chamber (MIC) based on miniature fission chamber mechanical parts, recently developed by the CEA (French Atomic Energy and Alternative Energies Commission) irradiated in the core of the Jožef Stefan Institute TRIGA Mark II reactor in Ljubljana, Slovenia. The aim of the study was to compare the measured MIC currents with calculated currents based on simulations with the MCNP6 code. A discrepancy of around 50% was observed between the measured and the calculated currents; in the latter taking into consideration only the prompt photon field. Further experimental measurements of MIC currents following reactor SCRAMs (reactor shutdown with rapid insertions of control rods) provide evidence that over 30% of the total measured signal is due to the delayed photon field, originating from fission and activation products, which are untreated in the calculations. In the comparison between the measured and calculated values, these findings imply an overall discrepancy of less than 20% of the total signal which is still unexplained.

  15. Testing WIMS-D4M cross sections and the ANL ENDF/B-V 69 group library. Results from global diffusion and Monte Carlo calculations compared with measurements in the Romanian 14-MW TRIGA reactor

    SciTech Connect

    Bretscher, M.M.

    1993-12-31

    The WIMS-D4 code has been modified (WIMS-D4M) to produce microscopic isotopic cross sections in ISOTXS format for use in diffusion and transport calculations. Beginning with 69-group libraries based on ENDF/B-V data, numerous cell calculations have been made to prepare a set of broad group cross sections for use in diffusion calculations. Global calculations have been made for two control rod states of the Romanian steady state TRIGA reactor with 29 fresh HEU fuel clusters. Detailed Monte Carlo calculations also have been performed for the same reactor configurations using data based on ENDF/B-V. Results from these global calculations are compared with each other and with the measured excess reactivities. Although region-averaged macroscopic principal cross sections obtained from WIMS-D4M are in good agreement with the corresponding Monte Carlo values, problems exist with the high energy (E > 10 keV) microscopic hydrogen transport cross sections.

  16. Design of neutron beams for neutron capture therapy using a 300-kW slab TRIGA reactor

    SciTech Connect

    Liu, H.B.

    1995-03-01

    A design for a slab reactor to produce an epithermal neutron beam and a thermal neutron beam for use in neutron capture therapy (NCT) is described. A thin reactor with two large-area faces, a ``slab`` reactor, was planned using eighty-six 20% enriched TRIGA fuel elements and four B{sub 4}C control rods. Two neutron beams were designed: an epithermal neutron beam from one face and a thermal neutron beam from the other. The planned facility, based on this slab-reactor core with a maximum operating power of 300 kW, will provide an epithermal neutron beam of 1.8 {times} 10{sup 9} n{sub epi}/cm{sup 2}{center_dot}s intensity with low contamination by fast neutrons and gamma rays and a thermal neutron beam of 9.0 {times} 10{sup 9}n{sub th}/cm{sup 2}{center_dot}s intensity with low fast-neutron dose and gamma dose. Both neutron beams will be forward directed. Each beam can be turned on and off independently through its individual shutter. A complete NCT treatment using the designed epithermal or thermal neutron beam would take 30 or 20 min, respectively, under the condition of assuming 10{mu}g {sup 10}B/g in the blood. Such exposure times should be sufficiently short to maintain near-optimal target (e.g., {sup 10}B, {sup 157}Gd, and {sup 235}U) distribution in tumor versus normal tissues throughout the irradiation. With a low operating power of 300 kW, the heat generated in the core can be removed by natural convection through a pool of light water. The proposed design in this study could be constructed for a dedicated clinical NCT facility that would operate very safely.

  17. Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System

    SciTech Connect

    Karim, Julia Abdul

    2008-05-20

    The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained.

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

    SciTech Connect

    Glumac, B; Ravnik, M.; Logar, M.

    1997-02-01

    Additional criticality safety analysis of a pool-type storage for TRIGA spent fuel at the Jozef Stefan Institute in Ljubljana, Slovenia, is presented. Previous results have shown that subcriticality is not guaranteed for some postulated accidents (earthquake with subsequent fuel rack disintegration resulting in contact fuel pitch) under the assumption that the fuel rack is loaded with fresh 12 wt% standard fuel. To mitigate this deficiency, a study was done on replacing a certain number of fuel elements in the rack with cadmium-loaded absorber rods. The Monte Carlo computer code MCNP4A with an ENDF/B-V library and detailed three-dimensional geometrical model of the spent-fuel rack was used for this purpose. First, a minimum critical number of fuel elements was determined for contact pitch, and two possible geometries of rack disintegration were considered. Next, it was shown that subcriticality can be ensured when pitch is decreased from a rack design pitch of 8 cm to contact, if a certain number of fuel elements (8 to 20 out of 70) are replaced by absorber rods, which are uniformly mixed into the lattice. To account for the possibility that random mixing of fuel elements and absorber rods can occur during rack disintegration and result in a supercritical configuration, a probabilistic study was made to sample the probability density functions for random absorber rod lattice loadings. Results of the calculations show that reasonably low probabilities for supercriticality can be achieved (down to 10{sup {minus}6} per severe earthquake, which would result in rack disintegration and subsequent maximum possible pitch decrease) even in the case where fresh 12 wt% standard TRIGA fuel would be stored in the spent-fuel pool.

  19. Automated system for neutron activation analysis determination of short lived isotopes at The DOW Chemical Company's TRIGA research reactor

    NASA Astrophysics Data System (ADS)

    Zieman, J. J.; Rigot, W. L.; Romick, J. D.; Quinn, T. J.; Kocher, C. W.

    1994-12-01

    An automated neutron activation analysis (NAA) system for the determination of short lived isotopes was constructed at The DOW Chemical Company's TRIGA Research Reactor in 1993. The NAA group of the Analytical Sciences Laboratory uses the reactor for thousands of analyses each year and therefore automation is important to achieve and maintain high throughput and precision (productivity). This project is complementary to automation of the long-lived counting facilities (see Romick et al., these Proceedings). Canberra/Nuclear Data Systems DEC-based software and electronics modules and an I/O mounting board are the basic commercial components. A Fortran program on a VAX computer controls I/O via ethernet to an Acquisition Interface Module (AIM). The AIM controls the γ spectrometer modules and is interfaced to a Remote Parallel Interface (RPI) module which controls the pneumatic transfer apparatus with TTL signals to the I/O mounting board. Near-infrared sensors are used to monitor key points in the transfer system. Spectra are acquired by a single HPGe detector mounted on a sliding rail to allow flexible and more reproducible counting geometries than with manual sample handling. The maximum sample size is 8 ml in a heat-sealed two dram vial. The sample vial is nested into a "rabbit" vial for irradiation which can be automatically removed prior to spectrum collection. The system was designed to be used by the reactor operator at the control console without the aid of an additional experimenter. Applications include the determination of selenium and silver in coal and water, fluorine in tetra-fluoro ethylene (TFE) coated membranes, aluminum and titanium in composite materials and trace fluorine in non-chlorinated cleaning solvents. Variable dead time software allows analysis for 77mSe despite high dead times from 16N encountered in samples.

  20. University Reactor Instrumentation Grant

    SciTech Connect

    S. M. Bajorek

    2000-02-01

    A noble gas air monitoring system was purchased through the University Reactor Instrumentation Grant Program. This monitor was installed in the Kansas State TRIGA reactor bay at a location near the top surface of the reactor pool according to recommendation by the supplier. This system is now functional and has been incorporated into the facility license.

  1. Comparison of EPR response of alanine and Gd₂O₃-alanine dosimeters exposed to TRIGA Mainz reactor.

    PubMed

    Marrale, M; Schmitz, T; Gallo, S; Hampel, G; Longo, A; Panzeca, S; Tranchina, L

    2015-12-01

    In this work we report some preliminary results regarding the analysis of electron paramagnetic resonance (EPR) response of alanine pellets and alanine pellets added with gadolinium used for dosimetry at the TRIGA research reactor in Mainz, Germany. Two set-ups were evaluated: irradiation inside PMMA phantom and irradiation inside boric acid phantom. We observed that the presence of Gd2O3 inside alanine pellets increases the EPR signal by a factor of 3.45 and 1.24 in case of PMMA and boric acid phantoms, respectively. We can conclude that in the case of neutron beam with a predominant thermal neutron component the addition of gadolinium oxide can significantly improve neutron sensitivity of alanine pellets. Monte Carlo (MC) simulations of both response of alanine and Gd-added alanine pellets with FLUKA code were performed and a good agreement was achieved for pure alanine dosimeters. For Gd2O3-alanine deviations between MC simulations and experimental data were observed and discussed. PMID:26315099

  2. Adaptation of triple axis neutron spectrometer for SANS measurements using alumina samples at TRIGA reactor of Bangladesh

    NASA Astrophysics Data System (ADS)

    Ahmed, F. U.; Kamal, I.; Yunus, S. M.; Datta, T. K.; Azad, A. K.; Zakaria, A. K. M.; Goyal, P. S.

    2005-09-01

    Double crystal method known as Bonse and Hart's technique has been employed to develop small angle neutron scattering (SANS) facility on a triple axis neutron spectrometer at TRIGA Mark II (3 MW) research reactor of Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. Two Si(1 1 1) crystals with very small mosaic spread ∼1 min have been used for this purpose. At an incident neutron wavelength of 1.24 Å, this device is useful for SANS in the Q range between 1.6×10 -3 and 10 -1 Å -1. This Q range allows investigating particle sizes and interparticle correlations on a length scale of ∼200 Å. Results of SANS experiments on three alumina (Al 2O 3) samples as performed using above setup are presented. It is seen that Al 2O 3 particles, indeed, scatter neutrons in regions of small angles. It is also seen that scattering is different for different samples showing that it changes with a change in particle size.

  3. Active in-core irradiation of SiC JFETs at 300 C in a TRIGA nuclear reactor

    SciTech Connect

    McGarrity, J.; Scozzie, C.; Blackburn, J.; DeLancey, M.

    1996-12-31

    In this paper the authors demonstrate that SiC transistors have the potential to operate in the severe high temperature and radiation environments of commercial and space nuclear power sources. 6H-SiC FETs were exposed to neutron fluxes and gamma dose rates as high as 1.6 {times} 10{sup 12} n/cm{sup 2}/sec and 3.8 {times} 10{sup 4} rad(Si)/sec while they were maintained under bias at both 300 C and room temperature within the core of a TRIGA reactor operated at 200 kW power level. The radiation exposure was continuous and the bias on the devices was interrupted only to record the current-voltage characteristics at various accumulated neutron fluences from 10{sup 13} to 5 {times} 10{sup 15} n/cm{sup 2}. No significant degradation in the device characteristics was observed until the total neutron fluence exceeded 10{sup 15} n/cm{sup 2} for irradiation at 25 C, and no significant changes were observed even at 5 {times} 10{sup 15} n/cm{sup 2} at 300 C.

  4. Assessment results of the Indonesian TRIGA SNF to be shipped to INEEL

    SciTech Connect

    Jefimoff, J.; Robb, A.K.; Wendt, K.M.; Syarip, I.; Alfa, T.

    1997-10-09

    This paper describes the Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) examination performed by technical personnel from the Idaho National Engineering and Environmental Laboratory (INEEL) at the Bandung and Yogyakarta research reactor facilities in Indonesia. The examination was required before the SNF would be accepted for transportation to and storage at the INEEL. This paper delineates the Initial Preparations prior to the Indonesian foreign research reactor (FRR) fuel examination. The technical basis for the examination, the TRIGA SNF Acceptance Criteria, and the physical condition required for transportation, receipt and storage of the TRIGA SNF at the INEEL is explained. In addition to the initial preparations, preparation descriptions of the Work Plan For TRIGA Fuel Examination, the Underwater Examination Equipment used, and personnel Examination Team Training are included. Finally, the Fuel Examination and Results of the aluminum and stainless steel clad TRIGA fuel examination have been summarized. Lessons learned from all the activities completed to date is provided in an addendum. The initial preparations included: (1) coordination between the INEEL, FRR or Badan Tenaga Atom Nasional (BATAN), DOE-HQ, and the US State Department and Embassy; (2) incorporating Savannah River Site (SRS) FRR experience and lessons learned; (3) collecting both FRR facility and spent fuel data, and issuing a radionuclide report (Radionuclide Mass Inventory, Activity, Decay Heat, and Dose Rate Parametric Data for TRIGA Spent Nuclear Fuels) needed for transportation and fuel acceptance at the INEEL; and (4) preexamination work at the research reactor for the fuel examination.

  5. 12. U.S. TRIGA users conference. Papers and abstracts

    SciTech Connect

    1990-07-01

    The Conference presentations were devoted to the following topics: new developments and improvements, including modifications of TRIGA reactors and equipment; experiments with TRIGA reactors (Neutron Radiography); radiochemistry, radioisotope production and beam irradiations (experiment applications, simulation); reactor physics - fuel utilization; reactor operation and maintenance experience; safety aspects, licensing and radiation protection.

  6. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    SciTech Connect

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

  7. The Design and Construction of a Cold Neutron Source for Use in the Cornell University Triga Reactor

    NASA Astrophysics Data System (ADS)

    Young, Lydia Jane

    A cold neutron source has been designed and constructed for insertion into the 6"-radial beam port of the Cornell University TRIGA reactor for use with a neutron guide tube system. The main differences between this cold source and other existing sources are the use of heat conduction as the method of cooling and the use of mesitylene (1,3,5 -trimethylbenzene; melting point, 228(DEGREES)K; boiling point, 437(DEGREES)K) as the moderating material. This thesis describes the design and construction details of the cold neutron source, discusses its safety aspects, and presents its cryogenic performance curves and also the results of a test of its neutron moderating ability. A closed-cycle helium gas refrigerator, located outside the reactor shielding, cools the 500 cm('3) moderator chamber and its surrounding heat shield by heat conduction through two meters of copper and rod tubing. Moderator temperatures of 23 (+OR-) 3(DEGREES)K have been achieved. Mesitylene, a hydrocarbon, is an effective cold moderator because even at low temperatures the weakly hindered rotational motions of its methyl groups enable the absorption of small amounts of energy ((LESSTHEQ) 0.005 eV) from neutrons. The use of mesitylene simplifies the cold source design because it is a liquid at room temperature and thus, the usual design safeguards required for sources using gaseous moderators are not necessary. Moreover, the flammability of mesitylene is much smaller than that of hydrogen and methane, which are the commonly used cold moderators. A method of transferring and handling the mesitylene, a carcinogen, was devised to ensure minimal contact with this substance. To test the neutron moderating ability of the cold neutron source, an out-of-reactor neutron transmission experiment was performed with the moderator chamber first at room temperature and then at about 23(DEGREES)K. The results indicate that the neutron energy spectrum is strongly shifted to lower energies when the chamber is cold

  8. 77 FR 42771 - License Renewal for the Dow Chemical TRIGA Research Reactor

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-07-20

    ... releases of radioactive effluents. As discussed in the NRC staff's safety evaluation, the systems and... that releases of radioactive material and personnel exposures were all well within applicable... Reactor Operations Gaseous radioactive effluents are discharged by the facility exhaust system via...

  9. Corrosion damage to the aluminum tank liner of the U.S. Geological Survey TRIGA Reactor

    SciTech Connect

    Perryman, R.E.; Millard, H.T. Jr.; Rusling, D.H.; Heifer, P.G.; Smith, W.L.

    1988-07-01

    During a routine maintenance small holes at the side of the tank of the reactor, penetrating the tank liner were discovered. Apparently the corrosion was acting from the back side of the tank forming the holes. The NRC was promptly notified and routine operations were suspended. Further investigation lead to the discovery of 74 holes, most of which were less than 1/8 inch in diameter with a few as large as 1/4 inch diameter. The results of an examination of the plate cut from the side of the tank correlated the absence of tar coating with the presence of numerous corrosion pits and craters. Along the welds in the corroded areas, parallel corrosion troughs existed on either side of the weld. Most of the pits and craters were too small to be detected by ultrasonic survey. In order to remedy the physical problem and be able to resume the reactor operation, a short-term strategy was adopted which involved covering the 74 holes with aluminum patches coated with epoxy. Reactor operations were resumed and over the next month four new holes were found and four patches applied. An inspection conducted after four months of operation found 28 new holes and the rate of leakage of water from the tank had increased to about 0.7 l/h. Because the rate of formation of holes seemed to be accelerating and the time required for maintenance was becoming unacceptable, it was decided to cease operation of the reactor until long-term repairs could be made. A new aluminum tank liner will be installed within the existing tank. A 2-inch wide annular void will then exist between the new and old liners. A pump will be installed inside the new liner to prevent the ground water from contacting it. The top of the void will be shielded to reduce the exposure to neutrons and gamma rays scattered from areas near the reactor. The reactor will be reinstalled at the bottom of the new liner on a plate which can be levelled from a distance of 10 feet.

  10. Neutron dosimetry and damage calculations for the TRIGA MARK-II reactor in Vienna

    NASA Astrophysics Data System (ADS)

    Weber, H. W.; Böck, H.; Unfried, E.; Greenwood, L. R.

    1986-02-01

    In order to improve the source characterization of the reactor, especially for recent irradiation experiments in the central irradiation thimble, neutron activation experiments were made on 16 nuclides and the neutron flux spectrum was adjusted using the computer code STAY'SL. The results for the total, thermal and fast neutron flux density at a reactor power of 250 kW are as follows: 2.1 × 10 17, 6.1 × 10 16 ( E < 0.55 eV), 7.6 × 10 16 ( E > 0.1 MeV) and 4.0 × 10 16 ( E > 1 MeV) m -2 s -1. respectively. Calculated damage energy cross sections and gas production rates are presented for selected elements.

  11. Material Sample Collection with Tritium and Gamma Analyses at the University of Illinois's Nuclear Research Laboratory TRIGA Nuclear Research Reactor

    SciTech Connect

    Charters, G.; Aggarwal, S.

    2006-07-01

    The University of Illinois in Champaign-Urbana has an Advanced TRIGA reactor facility which was built in 1960 and operated until August 1998. The facility was shutdown for a variety of reasons, primarily due to a lack of usage by the host institution. In 1998 the reactor went into SAFSTOR and finally shipped its fuel in 2004. At the present time a site characterization and decommissioning plan are in process and hope to be submitted to the NRC in early 2006. The facility had to be fully characterized and part of this characterization involved the collection and analysis of samples. This included various solid media such as, concrete, graphite, metals, and sub-slab surface soils for immediate analysis of Activation and Tritium contamination well below the easily measured surfaces. This detailed facility investigation provided a case to eliminate historical unknowns, increasing the confidence for the segregation and packaging of high specific activity Low Level Radwaste (LLRW), from which a strategy of 'surgical-demolition' and segregation could be derived thus maximizing the volumes of 'clean material'. Performing quantitative volumetric concrete or metal radio-analyses safer and faster (without lab intervention) was a key objective of this dynamic characterization approach. Currently, concrete core bores are shipped to certified laboratories where the concrete residue is run through a battery of tests to determine the contaminants. The existing core boring operation volatilises or washes out some of the contaminants (like tritium) and oftentimes cross-contaminates the are a around the core bore site. The volatilization of the contaminants can lead to airborne problems in the immediate vicinity of the core bore. Cross-contamination can increase the contamination area and thereby increase the amount of waste generated that needs to be treated and stabilized before disposal. The goal was to avoid those field activities that could cause this type of release. Therefore

  12. Criticality and Safety Parameter Studies of a 3-MW TRIGA MARK-II Research Reactor and Validation of the Generated Cross-Section Library and Computational Method

    SciTech Connect

    Bhuiyan, S.I.; Mondal, M.A.W.; Sarker, M.M.; Rahman, M.; Shahdatullah, M.S.; Huda, M.Q.; Chakrobortty, T.K.; Khan, M.J.H

    2000-05-15

    are specially treated with the Prize options available in WIMSD-5 to take into account the fine structure of the flux gradient in the fuel-reflector interface region. The hot spot is found physically at the fuel position C4 with a maximum power density of 1.044559 x 10{sup 2} W/cm{sup 3}. The calculated total peaking factor is 5.8867 compared to the original SAR value of 5.6325. The curve of {alpha}{sub f} with the temperature at zero burnup shows that the curve deviates somewhat with that reported in the original SAR for low-enriched uranium fuel. The MCNP calculations establish that the CITATION calculations and the generated cross-section library are reasonably good for neutronic analysis of TRIGA reactors. The results obtained from the neutronic analysis will be used to analyze the thermal-hydraulic behavior and the safety margins of the core both for steady-state and pulse-mode operations.

  13. Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

    SciTech Connect

    Maskin, Mazleha; Tom, Phongsakorn Prak; Lanyau, Tonny Anak; Saad, Mohamad Fauzi; Ismail, Ahmad Razali; Abu, Mohamad Puad Haji; Brayon, Fedrick Charlie Matthew; Mohamed, Faizal

    2014-02-12

    As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia.

  14. Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

    NASA Astrophysics Data System (ADS)

    Maskin, Mazleha; Tom, Phongsakorn Prak; Lanyau, Tonny Anak; Brayon, Fedrick Charlie Matthew; Mohamed, Faizal; Saad, Mohamad Fauzi; Ismail, Ahmad Razali; Abu, Mohamad Puad Haji

    2014-02-01

    As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia.

  15. Methods of reducing liquid effluent from the OSU TRIGA MKII Reactor

    SciTech Connect

    Higginbotham, J.F.; Dodd, B.; Pratt, D.S.; Smith, S.; Anderson, T.V.

    1992-07-01

    In 1991, the OSU Radiation Center implemented a program to minimize the liquid effluent generated by the reactor facility. The goal of program is to become a 'zero' release facility with regards to routine liquid discharges. Only two liquid waste streams exist for the OSU reactor facility: discharges resulting from changing resin in the deminerializer and decontamination of equipment, primarily sample loading tubes. This paper describes a system which allows remote resin exchange to performed with the collection of all flush water. This water is then recycled for use as makeup for the primary water system. The service life of the resin is maximized by using a steam distillation unit as the source of makeup water to the deminerializer system instead of water coming directly from the City of Corvallis water supply. The second source of liquid waste water comes from the decontamination of the plastic loading tubes used to encapsulate samples. This process originally involved placing the tubes in a dishwasher and sending the discharge to a hold up tank. If the radionuclide concentrations in the tank were below the maximum permissible concentrations of 10CFR20 then it was released to the sanitary sewerage. This process was replaced in 1991 with a system which involved manual washing and rinsing of the tubes with the liquids being absorbed for disposal as solid waste. This paper will also describe the system which is being built to replace this process. It will use the dishwasher unit again but the liquid discharge will collected for absorption and disposal as solid waste. (author)

  16. Preliminary TRIGA fuel burn-up evaluation by means of Monte Carlo code and computation based on total energy released during reactor operation

    SciTech Connect

    Borio Di Tigliole, A.; Bruni, J.; Panza, F.; Alloni, D.; Cagnazzo, M.; Magrotti, G.; Manera, S.; Prata, M.; Salvini, A.; Chiesa, D.; Clemenza, M.; Pattavina, L.; Previtali, E.; Sisti, M.; Cammi, A.

    2012-07-01

    Aim of this work was to perform a rough preliminary evaluation of the burn-up of the fuel of TRIGA Mark II research reactor of the Applied Nuclear Energy Laboratory (LENA) of the Univ. of Pavia. In order to achieve this goal a computation of the neutron flux density in each fuel element was performed by means of Monte Carlo code MCNP (Version 4C). The results of the simulations were used to calculate the effective cross sections (fission and capture) inside fuel and, at the end, to evaluate the burn-up and the uranium consumption in each fuel element. The evaluation, showed a fair agreement with the computation for fuel burn-up based on the total energy released during reactor operation. (authors)

  17. GEANT4 used for neutron beam design of a neutron imaging facility at TRIGA reactor in Morocco

    NASA Astrophysics Data System (ADS)

    Ouardi, A.; Machmach, A.; Alami, R.; Bensitel, A.; Hommada, A.

    2011-09-01

    Neutron imaging has a broad scope of applications and has played a pivotal role in visualizing and quantifying hydrogenous masses in metallic matrices. The field continues to expand into new applications with the installation of new neutron imaging facilities. In this scope, a neutron imaging facility for computed tomography and real-time neutron radiography is currently being developed around 2.0MW TRIGA MARK-II reactor at Maamora Nuclear Research Center in Morocco (Reuscher et al., 1990 [1]; de Menezes et al., 2003 [2]; Deinert et al., 2005 [3]). The neutron imaging facility consists of neutron collimator, real-time neutron imaging system and imaging process systems. In order to reduce the gamma-ray content in the neutron beam, the tangential channel was selected. For power of 250 kW, the corresponding thermal neutron flux measured at the inlet of the tangential channel is around 3×10 11 ncm 2/s. This facility will be based on a conical neutron collimator with two circular diaphragms with diameters of 4 and 2 cm corresponding to L/D-ratio of 165 and 325, respectively. These diaphragms' sizes allow reaching a compromise between good flux and efficient L/D-ratio. Convergent-divergent collimator geometry has been adopted. The beam line consists of a gamma filter, fast neutrons filter, neutron moderator, neutron and gamma shutters, biological shielding around the collimator and several stages of neutron collimator. Monte Carlo calculations by a fully 3D numerical code GEANT4 were used to design the neutron beam line ( http://www.info.cern.ch/asd/geant4/geant4.html[4]). To enhance the neutron thermal beam in terms of quality, several materials, mainly bismuth (Bi) and sapphire (Al 2O 3) were examined as gamma and neutron filters respectively. The GEANT4 simulations showed that the gamma and epithermal and fast neutron could be filtered using the bismuth (Bi) and sapphire (Al 2O 3) filters, respectively. To get a good cadmium ratio, GEANT 4 simulations were used to

  18. Neutron and gamma radiography of UO{sub 2} and TRIGA fuel elements

    SciTech Connect

    Robinson, A.H.; Gao, Y.C.; Johnson, A.G.; Ringle, J.C.

    1982-07-01

    The Oregon State TRIGA Reactor neutron radiography facility has been used to produce both neutron and gamma radiographs of reactor fuel. In this paper a comparison of the applicability of neutron and gamma radiography to both UO{sub 2} fuel pins and TRIGA fuel elements is made. In the case of UO{sub 2} fuel, conventional thermal neutron radiography produces excellent quality radiographs. These radiographs may be used to detect various defects in the fuel such as enrichment differences, cracks, end-capping, inclusions, etc. For TRIGA fuel elements, conventional thermal neutron radiography will not show the internal structure. This is due to the high hydrogen content of the fuel. These elements are typically 8.5 w/o uranium in Zr-H{sub 1.7}; the density of hydrogen in the fuel being about 80 percent that of water. Further, while epithermal radiography significantly improves the radiographs, defects may go undetected. As an alternative to neutron radiography, high energy gamma radiographs of TRIGA fuel elements have been taken using the same facility. The gamma spectrum emitted by the reactor core is sufficiently high in energy that very good radiographs may be obtained with this technique. These radiographs show excellent detail for the internal structure of the TRIGA fuel. (author)

  19. Corrosion in the aluminum containment tank at the Nuclear Center of Mexico TRIGA Mark III reactor

    SciTech Connect

    Mota, Juan Ramon

    1986-07-01

    The reactor developed a leak inside the exposure room discovered when it was opened for a routine inspection. This leak started to diminish immediately after it was found and disappeared completely in 2.5 months. The hydrostatic tests of the exposure room cooling water pipes and of the primary cooling system suction pipe proved that piping do not have leaks. A portion of the total volume of water was drained from the pool to conduct an inspection on the aluminum liner. Penetrant dye tests were initiated over welded Joints and walls. Welded Joints were all found to be in good condition but a total of 35 indications were reported on walls and concentrated on two main areas. A vacuum system was used to test for leakage. Seven indications were found to be perforations that crossed through the wall, fifteen indications did not cross through the wall but required repair and the rest were superficial irregularities. For the inspection of surfaces that remained covered by water, two methods were used. One was a television camera that was adapted to be used under water and hooked to a monitor and a videorecorder for close up inspection of the walls. The other consisted of submarine still color photography performed by divers. The evaluation of these inspections concluded that out of the 10 areas previously identified, only one presented the kind of problem that required repair. The last inspection performed was that using ultrasound techniques. Irregularities found did not require complete replacement of the aluminum liner. The repair procedures included the welding of aluminum plates over damaged areas and the injection of an effective insulating material (resin) to stop the corrosion mechanism.

  20. 7. biennial U.S. TRIGA users' conference. Papers and abstracts

    SciTech Connect

    1980-07-01

    The conference covers the following topics: new developments in the TRIGA system; uses of microprocessors in control and monitoring and measurement of TRIGA performance parameters; safeguards, emergency planning, reactor standards; research facilities, fuel tests and calculations; TRIGA reactor parameters: emergency training.

  1. Solid State Reactor Final Report

    SciTech Connect

    Mays, G.T.

    2004-03-10

    The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas of research

  2. Validation of absolute axial neutron flux distribution calculations with MCNP with 197Au(n,γ)198Au reaction rate distribution measurements at the JSI TRIGA Mark II reactor.

    PubMed

    Radulović, Vladimir; Štancar, Žiga; Snoj, Luka; Trkov, Andrej

    2014-02-01

    The calculation of axial neutron flux distributions with the MCNP code at the JSI TRIGA Mark II reactor has been validated with experimental measurements of the (197)Au(n,γ)(198)Au reaction rate. The calculated absolute reaction rate values, scaled according to the reactor power and corrected for the flux redistribution effect, are in good agreement with the experimental results. The effect of different cross-section libraries on the calculations has been investigated and shown to be minor. PMID:24316530

  3. Neutron activation analyses and half-life measurements at the usgs triga reactor

    NASA Astrophysics Data System (ADS)

    Larson, Robert E.

    Neutron activation of materials followed by gamma spectroscopy using high-purity germanium detectors is an effective method for making measurements of nuclear beta decay half-lives and for detecting trace amounts of elements present in materials. This research explores applications of neutron activation analysis (NAA) in two parts. Part 1. High Precision Methods for Measuring Decay Half-Lives, Chapters 1 through 8 Part one develops research methods and data analysis techniques for making high precision measurements of nuclear beta decay half-lives. The change in the electron capture half-life of 51Cr in pure chromium versus chromium mixed in a gold lattice structure is explored, and the 97Ru electron capture decay half-life are compared for ruthenium in a pure crystal versus ruthenium in a rutile oxide state, RuO2. In addition, the beta-minus decay half-life of 71mZn is measured and compared with new high precision findings. Density Functional Theory is used to explain the measured magnitude of changes in electron capture half-life from changes in the surrounding lattice electron configuration. Part 2. Debris Collection Nuclear Diagnostic at the National Ignition Facility, Chapters 9 through 11 Part two explores the design and development of a solid debris collector for use as a diagnostic tool at the National Ignition Facility (NIF). NAA measurements are performed on NIF post-shot debris collected on witness plates in the NIF chamber. In this application NAA is used to detect and quantify the amount of trace amounts of gold from the hohlraum and germanium from the pellet present in the debris collected after a NIF shot. The design of a solid debris collector based on material x-ray ablation properties is given, and calculations are done to predict performance and results for the collection and measurements of trace amounts of gold and germanium from dissociated hohlraum debris.

  4. Measurement of DNA damage induced by irradiation with gamma-rays from a TRIGA Mark II research reactor in human cells using Fast Micromethod.

    PubMed

    Hassanein, Hamdy; Müller, Claudia I; Schlösser, Dietmar; Kratz, Karl-Ludwig; Senyuk, Olga F; Schröder, Heinz C

    2002-06-01

    The Fast Micromethod is a novel quick and convenient microplate assay for determination of DNA single-strand breaks. This method measures the rate of unwinding of cellular DNA upon exposure to alkaline conditions using a fluorescent dye which preferentially binds to double-stranded DNA. Here we applied this method to determine the levels of DNA single-strand breaks in HeLa cells induced by y-irradiation deriving from fission isotopes and activation products at the TRIGA Mark II research reactor in Mainz. An increased strand scission factor (SSF) value, which is indicative for DNA damage, was found at doses of 1 Gy and higher. A similar increase in SSF value, which further increased in a dose-dependent manner, was found in human peripheral blood mononuclear cells after irradiation with 6 MV X-rays from a linear accelerator to give a total exposure of 0.5 to 10 Gy. PMID:12064446

  5. Chromosome aberrations induced in human lymphocytes by U-235 fission neutrons: I. Irradiation of human blood samples in the "dry cell" of the TRIGA Mark II nuclear reactor.

    PubMed

    Fajgelj, A; Lakoski, A; Horvat, D; Remec, I; Skrk, J; Stegnar, P

    1991-11-01

    A set-up for irradiation of biological samples in the TRIGA Mark II research reactor in Ljubljana is described. Threshold activation detectors were used for characterisation of the neutron flux, and the accompanying gamma dose was measured by TLDs. Human peripheral blood samples were irradiated "in vitro" and biological effects evaluated according to the unstable chromosomal aberrations induced. Biological effects of two types of cultivation of irradiated blood samples, the first immediately after irradiation and the second after 96 h storage, were studied. A significant difference in the incidence of chromosomal aberrations between these two types of samples was obtained, while our dose-response curve fitting coefficients alpha 1 = (7.71 +/- 0.09) x 10(-2) Gy-1 (immediate cultivation) and alpha 2 = (11.03 +/- 0.08) x 10(-2) Gy-1 (96 h delayed cultivation) are in both cases lower than could be found in the literature. PMID:1962281

  6. Dose calculation in biological samples in a mixed neutron-gamma field at the TRIGA reactor of the University of Mainz.

    PubMed

    Schmitz, Tobias; Blaickner, Matthias; Schütz, Christian; Wiehl, Norbert; Kratz, Jens V; Bassler, Niels; Holzscheiter, Michael H; Palmans, Hugo; Sharpe, Peter; Otto, Gerd; Hampel, Gabriele

    2010-10-01

    To establish Boron Neutron Capture Therapy (BNCT) for non-resectable liver metastases and for in vitro experiments at the TRIGA Mark II reactor at the University of Mainz, Germany, it is necessary to have a reliable dose monitoring system. The in vitro experiments are used to determine the relative biological effectiveness (RBE) of liver and cancer cells in our mixed neutron and gamma field. We work with alanine detectors in combination with Monte Carlo simulations, where we can measure and characterize the dose. To verify our calculations we perform neutron flux measurements using gold foil activation and pin-diodes. Material and methods. When L-α-alanine is irradiated with ionizing radiation, it forms a stable radical which can be detected by electron spin resonance (ESR) spectroscopy. The value of the ESR signal correlates to the amount of absorbed dose. The dose for each pellet is calculated using FLUKA, a multipurpose Monte Carlo transport code. The pin-diode is augmented by a lithium fluoride foil. This foil converts the neutrons into alpha and tritium particles which are products of the (7)Li(n,α)(3)H-reaction. These particles are detected by the diode and their amount correlates to the neutron fluence directly. Results and discussion. Gold foil activation and the pin-diode are reliable fluence measurement systems for the TRIGA reactor, Mainz. Alanine dosimetry of the photon field and charged particle field from secondary reactions can in principle be carried out in combination with MC-calculations for mixed radiation fields and the Hansen & Olsen alanine detector response model. With the acquired data about the background dose and charged particle spectrum, and with the acquired information of the neutron flux, we are capable of calculating the dose to the tissue. Conclusion. Monte Carlo simulation of the mixed neutron and gamma field of the TRIGA Mainz is possible in order to characterize the neutron behavior in the thermal column. Currently we also

  7. Behavior of 12 wt% TRIGA fuel after many years of operation

    SciTech Connect

    Levine, S.H.; Boyle, P.

    1997-12-01

    In July 1972, six 12 wt% Uzr-H TRIGA fuel elements were loaded into the B-ring, the innermost ring, of the Penn State Breazeale Research Reactor (PSBR) to increase its k{sub eff}. Of these initial six fuel elements, three remain in the core, and the other three fuel elements had to be removed from the core. The purpose of this summary is to present operational-type data on the 12 wt% Uzr-H TRIGA fuel elements that have been in the PSBR during the past 25 yr and to postulate reasons for the structural change of those removed from the core.

  8. Operational characteristics of the new 12 wt% TRIGA fuel

    SciTech Connect

    Boyle, P.; Levine, S.H.

    1997-12-01

    It has been reported that an instrumented TRIGA fuel element, I-15, had higher than normal fuel temperatures. As shown in Fig. 1, one of the newest instrumented fuel elements, I-17, has fuel temperatures equal to/and or higher than that of I-15. Another new fuel element, I-16, behaves similarly to I-17, and this is before they have been pulsed (the first pulses increase the steady-state temperature). Thus, there is a significant increase in the measured fuel temperatures of the newest TRIGA fuel received from General Atomics. They are estimated to measure fuel temperatures that exceed 600{degrees}C when in the B-ring at 1 MW at their beginning of life (BOL). The purpose of this summary is to report on the measurements performed with these new fuel elements and to describe the effect that the new information is having on the future fuel management plans for the Penn State Breazeale Research Reactor.

  9. Dose estimation in B16 tumour bearing mice for future irradiation in the thermal column of the TRIGA reactor after B/Gd/LDL adduct infusion.

    PubMed

    Protti, N; Ballarini, F; Bortolussi, S; Bruschi, P; Stella, S; Geninatti, S; Alberti, D; Aime, S; Altieri, S

    2011-12-01

    To test the efficacy of a new (10)B-vector compound, the B/Gd/LDL adduct synthesised at Torino University, in vivo irradiations of murine tumours are in progress at the TRIGA Mark II reactor of the Pavia University. A localised B16 melanoma tumour is generated in C57BL/6 mice and subsequently infused with the adduct. During the irradiation, the mouse will be put in a shield to protect the whole body except the tumour in the back-neck area. To optimise the treatment set-up, MCNP simulations were performed. A very simplified mouse model was built using MCNP geometry capabilities, as well as the geometry of the shield made of 99% (10)B enriched boric acid. A hole in the shield is foreseen in correspondence of the back-neck region. Many configurations of the shield were tested in terms of neutron flux, dose distribution and mean induced activity in the tumour region and in the radiosensitive organs of the mouse. In the final set-up, up to five mice can be treated simultaneously in the reactor thermal column and the neutron fluence in the tumour region for 10 min of irradiation is of about 5×10(12) cm(-2). PMID:21459587

  10. Calculations of dose distributions in the lungs of a rat model irradiated in the thermal column of the TRIGA reactor in Pavia.

    PubMed

    Protti, N; Bortolussi, S; Stella, S; Gadan, M A; De Bari, A; Ballarini, F; Bruschi, P; Ferrari, C; Clerici, A M; Zonta, C; Bakeine, J G; Dionigi, P; Zonta, A; Altieri, S

    2009-07-01

    To test the possibility to apply boron neutron capture therapy (BNCT) to lung tumors, some rats are planned to be irradiated in the thermal column of the TRIGA reactor of the University of Pavia. Before the irradiation, lung metastases will be induced in BDIX rats, which will be subsequently infused with boronophenylalanine (BPA). During the irradiation, the rats will be positioned in a box designed to shield the whole animal except the thorax area. In order to optimize the irradiation set-up and to design a suitable shielding box, a set of calculations were performed with the MCNP Monte Carlo transport code. A rat model was constructed using the MCNP geometry capabilities and was positioned in a box with walls filled with lithium carbonate. A window was opened in front of the lung region. Different shapes of the holder and of the window were tested and analyzed in terms of the dose distribution obtained in the lungs and of the dose absorbed by the radiosensitive organs in the rat. The best configuration of the holder ensures an almost uniform thermal neutron flux inside the lungs (Phi(max)/Phi(min)=1.5), an irradiation time about 10 min long, to deliver at least 40 Gy(w) to the tumor, a mean lung dose of 5.9+/-0.4 Gy(w), and doses absorbed by all the other healthy tissues below the tolerance limits. PMID:19406647

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

    SciTech Connect

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

    1994-12-31

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

  12. TRIGA Mark II benchmark experiment; Part II: Pulse operation

    SciTech Connect

    Mele, I.; Ravnik, M.; Trkov, A. )

    1994-01-01

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

  13. Thermal hydraulic calculations to support increase in operating power in McClellen Nuclear Radiation Center(MNRC) TRIGA reactor.

    SciTech Connect

    Jensen, R. T.

    1998-05-05

    The RELAP5/Mod3.1 computer program has been used to successfully perform thermal-hydraulic analyses to support the Safety Analysis for increasing the MNRC reactor from 1.0 MW to 2.0 MW. The calculation results show the reactor to have operating margin for both the fuel temperature and critical heat flux limits. The calculated maximum fuel temperature of 705 C is well below the 750 C operating limit. The critical heat flux ratio was calculated to be 2.51.

  14. Feasibility study of the University of Utah TRIGA reactor power upgrade in respect to control rod system

    NASA Astrophysics Data System (ADS)

    Cutic, Avdo

    The objectives of this thesis are twofold: to determine the highest achievable power levels of the current University of Utah TRIG Reactor (UUTR) core configuration with the existing three control rods, and to design the core for higher reactor power by optimizing the control rod worth. For the current core configuration, the maximum reactor power, eigenvalue keff, shutdown margin, and excess reactivity have been measured and calculated. These calculated estimates resulted from thermal power calibrations, and the control rod worth measurements at various power levels. The results were then used as a benchmark to verify the MCNP5 core simulations for the current core and then to design a core for higher reactor power. This study showed that the maximum achievable power with the current core configuration and control rod system is 150kW, which is 50kW higher than the licensed power of the UUTR. The maximum achievable UUTR core power with the existing fuel is determined by optimizing the core configuration and control rod worth, showing that a power upgrade of 500 kW is achievable. However, it requires a new control rod system consisting of a total of four control rods. The cost of such an upgrade is $115,000.

  15. Chemical and material studies to understand the source of corrosion in the Geological Survey TRIGA Reactor (GSTR) tank liner

    SciTech Connect

    Rusling, D.H.; Millard, H.T. Jr.; Heifer, P.G.; Perryman, R.E.; Smith, W.L.

    1988-07-01

    Corrosion damage to the aluminum tank liner of the GSTR reactor was discovered and samples of various materials were collected for chemical and mineralogical analyses. The following scenario for the corrosion was suggested: 1. Cyclical temperature changes caused the tank liner to change size repeatedly. It extruded tar as it expanded and created voids as it contracted. 2. Hydrostatic pressure forced ground water through openings in the concrete into voids near the bottom of the tank, and overflow introduced tank water at the top of the tank. 3. The expansion-contraction cycle moved the water around the complex, interconnecting systems of voids and, in some locations, caused the tar-to-aluminum bond to fail. 4. Chemical interactions of the water with the tar and concrete supplied the elements capable of corroding the aluminum (e.g., Zn, Cu). 5. The corrosive solution has reacted with the aluminum over the lifetime of the reactor to produce the present corrosion damage. 6. As corrosion pits became holes, reactor tank water entered the voids.

  16. Radioactivity of spent TRIGA fuel

    NASA Astrophysics Data System (ADS)

    Usang, M. D.; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

    2015-04-01

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

  17. Radioactivity of spent TRIGA fuel

    SciTech Connect

    Usang, M. D. Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

    2015-04-29

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

  18. The U-ZrH{sub x} alloy: Its properties and use in TRIGA fuel

    SciTech Connect

    Simnad, M.T.

    1980-07-01

    The uranium-zirconium-hydride fuel is an integral fuel-moderator system. Development of the UZr-hydride fuel technology has been under way at General Atomic since 1957. During this period over 6000 fuel elements have been fabricated for the TRIGA reactors. Over 25,000 pulses have been performed with the TRIGA fuel elements at General Atomic. The TRIGA fuel was developed around the concept of inherent safety. The development and the characteristics of the TRIGA fuels are described in this paper. The fabrication techniques have been developed to the point where the production of fuel bodies containing controlled amounts of hydrogen and burnable poison (erbium) has been carried out in sizes up to 1.5 in. in diameter. The instrumented fuel elements have been designed to determine the temperatures in the fuels and claddings and to record the gas pressures in the fuel elements, under both steady- state and pulsed operations. The physical, mechanical and corrosion properties of the fuel are presented, along with empirical correlations relating irradiation behavior and fission product retention to temperature, composition, burnup, and neutron flux and fluence. (author)

  19. Northrop Triga facility decommissioning plan versus actual results

    SciTech Connect

    Gardner, F.W.

    1986-01-01

    This paper compares the Triga facility decontamination and decommissioning plan to the actual results and discusses key areas where operational activities were impacted upon by the final US Nuclear Regulatory Commission (NRC)-approved decontamination and decommissioning plan. Total exposures for fuel transfer were a factor of 4 less than planned. The design of the Triga reactor components allowed the majority of the components to be unconditionally released.

  20. 77 FR 68155 - The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R-84

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-11-15

    ... NRC's E-Filing rule (72 FR 49139; August 28, 2007). The E-Filing process requires participants to... filing requirements of the NRC's E-Filing Rule (72 FR 49139; August 28, 2007) apply to appeals of NRC... at a maximum steady-state thermal power of 1.1 megawatts (MW). The renewed license would...

  1. Steady-state spheromak reactor studies. Revision

    SciTech Connect

    Krakowski, R.A.; Hagenson, R.L.

    1985-01-01

    After summarizing the essential elements of a gun-sustained spheromak, the potential for a steady-state is explored by means of a comprehensive physics/engineering/costing model. A range of cost-optimized reactor design point is presented, and the sensitivity of cost to key physics, engineering, and operational variables is reported.

  2. Status of the TRIGA shipments to the INEEL from Europe

    SciTech Connect

    Mustin, T.; Stump, R.C.; Tyacke, M.J.

    1997-10-09

    This paper reports the activities underway by the US Department of Energy (DOE) for returning Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) from foreign research reactors (FRR) in four European countries to the Idaho National Engineering and Environmental Laboratory (INEEL). Those countries are Germany, Italy, Romania, and Slovenia. This is part of the ``Nuclear Weapons Nonproliferation Policy`` of returning research reactor SNF containing uranium enriched in the US. This paper describes the results of a pre-assessment trip in September, 1997, to these countries, including: history of the reactors and research being performed; inventory of TRIGA SNF; fuel types (stainless steel, aluminum, or Incoloy) and enrichments; and each country`s plans for returning their TRIGA SNF to the INEEL.

  3. An RFQ cooler and buncher for the TRIGA-SPEC experiment

    NASA Astrophysics Data System (ADS)

    Beyer, T.; Blaum, K.; Block, M.; Düllmann, Ch. E.; Eberhardt, K.; Eibach, M.; Frömmgen, N.; Geppert, C.; Gorges, C.; Grund, J.; Hammen, M.; Kaufmann, S.; Krieger, A.; Nagy, Sz.; Nörterhäuser, W.; Renisch, D.; Smorra, C.; Will, E.

    2014-01-01

    A linear Paul trap for cooling of ion beams, the former cooler for emittance elimination radiofrequency quadrupole (RFQ) at MISTRAL/ISOLDE, has been installed and commissioned at the TRIGA-SPEC experiment located at the research reactor TRIGA Mainz. It is connected to a hot-surface-ionization ion source and a subsequent mass separator for ionization and pre-separation of neutron-rich fission products as delivered from the reactor. The capability of accumulating and bunching ion beams has been implemented to provide low-emittance ion pulses of 250 ns width containing up to 106 ions. A technical description of the upgraded RFQ as well as its characterization with stable ions is presented. Its installation allows delivery of low-emittance ion bunches to the two branches of the TRIGA-SPEC experiment, namely TRIGA-TRAP and TRIGA-LASER.

  4. 10. biennial U.S. TRIGA users' conference. Papers and abstracts

    SciTech Connect

    1986-07-01

    The conference cover the following main topics for TRIGA reactors: reactor instrumentation and measurements of reactor parameters, reactor operation and modifications, design innovation and service works, fast neutron spectrum, fuel examination, neutron flux, heat transfer, accidents analysis, corrosion problems, fuel failures and fuel management, mechanical problems and maintenance.

  5. Hybrid Reactor Simulation of Boiling Water Reactor Power Oscillations

    SciTech Connect

    Huang Zhengyu; Edwards, Robert M.

    2003-08-15

    Hybrid reactor simulation (HRS) of boiling water reactor (BWR) instabilities, including in-phase and out-of-phase (OOP) oscillations, has been implemented on The Pennsylvania State University TRIGA reactor. The TRIGA reactor's power response is used to simulate reactor neutron dynamics for in-phase oscillation or the fundamental mode of the reactor modal kinetics for OOP oscillations. The reactor power signal drives a real-time boiling channel simulation, and the calculated reactivity feedback is in turn fed into the TRIGA reactor via an experimental changeable reactivity device. The thermal-hydraulic dynamics, together with first harmonic mode power dynamics, is digitally simulated in the real-time environment. The real-time digital simulation of boiling channel thermal hydraulics is performed by solving constitutive equations for different regions in the channel and is realized by a high-performance personal computer. The nonlinearity of the thermal-hydraulic model ensures the capability to simulate the oscillation phenomena, limit cycle and OOP oscillation, in BWR nuclear power plants. By adjusting reactivity feedback gains for both modes, various oscillation combinations can be realized in the experiment. The dynamics of axially lumped power distribution over the core is displayed in three-dimensional graphs. The HRS reactor power response mimics the BWR core-wide power stability phenomena. In the OOP oscillation HRS, the combination of reactor response and the simulated first harmonic power using shaping functions mimics BWR regional power oscillations. With this HRS testbed, a monitoring and/or control system designed for BWR power oscillations can be experimentally tested and verified.

  6. Solid State Molecular Reactors in Space

    NASA Astrophysics Data System (ADS)

    Maurette, M.

    2011-03-01

    Lunar minerals and impact glasses, convert the polyatomic beam of solar wind (SW) ions into a flux of small molecules (e.g., H2, N2, H2O, CO, CO2, CH4, C2H4, C2H6, HCN, metal carbides and deuterides, etc.). They thus behave as "Solid State Molecular Reactors". Moreover, ~100-200 μm size micrometeoroids (μMs) have also been exposed to the SW in the zodiacal cloud, before being captured by the Earth and recovered as Antarctic micrometeorites. They are mostly composed of a PAH-rich hydrous-carbonaceous material, which amplifies their power as molecular reactors. In particular, during the first ~200 Myr of the post-lunar period, about 75% of the μMs have been melted and/or volatilized upon atmospheric entry. The release of their volatile species triggered a cosmic volcanism around the mesopause that ruled the formation of the early Earth's atmosphere and climate. Furthermore, a fraction of the μMs that survive unmelted upon atmospheric entry did settle on the proto-oceans floors. Upon further burial in sediments their constituent PAH-rich kerogen was cracked into abiotic oil, which generated giant oil slicks that fed prebiotic chemistry. Many stars, of all ages and types, are embedded into a secondary debris-disk loaded with ion implanted μMs. Some of them are expelled to the interstellar medium (ISM) where they behave first as "dormant-invisible" molecular reactors, until they became reactivated by various processes to synthesize interstellar molecules. This short paper only focus on some highlights of this research dealing with the synthesis of important interstellar molecules, including the most abundant ones (H2 and CO) and H2O, HCN and PAHs, all involved in prebiotic chemistry.

  7. Hot Fuel Examination Facility's neutron radiography reactor

    SciTech Connect

    Pruett, D.P.; Richards, W.J.; Heidel, C.C.

    1983-01-01

    Argonne National Laboratory-West is located near Idaho Falls, Idaho, and is operated by the University of Chicago for the United States Department of Energy in support of the Liquid Metal Fast Breeder Reactor Program, LMFBR. The Hot Fuel Examination Facility, HFEF, is one of several facilities located at the Argonne Site. HFEF comprises a large hot cell where both nondestructive and destructive examination of highly-irradiated reactor fuels are conducted in support of the LMFBR program. One of the nondestructive examination techniques utilized at HFEF is neutron radiography, which is provided by the NRAD reactor facility (a TRIGA type reactor) below the HFEF hot cell.

  8. Research reactors - an overview

    SciTech Connect

    West, C.D.

    1997-03-01

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

  9. Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

    SciTech Connect

    Edwards, Robert; Huang, Zhengyu

    2001-06-17

    The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown.

  10. Transport of fission products with a helium gas-jet at TRIGA-SPEC

    NASA Astrophysics Data System (ADS)

    Eibach, M.; Beyer, T.; Blaum, K.; Block, M.; Eberhardt, K.; Herfurth, F.; Geppert, C.; Ketelaer, J.; Ketter, J.; Krämer, J.; Krieger, A.; Knuth, K.; Nagy, Sz.; Nörtershäuser, W.; Smorra, C.

    2010-02-01

    A helium gas-jet system for the transport of fission products from the research reactor TRIGA Mainz has been developed, characterized and tested within the TRIGA-SPEC experiment. For the first time at TRIGA Mainz carbon aerosol particles have been used for the transport of radionuclides from a target chamber with high efficiency. The radionuclides have been identified by means of γ-spectroscopy. Transport time, efficiency as well as the absolute number of transported radionuclides for several species have been determined. The design and the characterization of the gas-jet system are described and discussed.

  11. Interpretation of TRIGA reactivity transients with RELAP5/PARCS coupled-code

    SciTech Connect

    Bandini, G.; Meloni, P.; Polidori, M.

    2006-07-01

    In the frame of future experiments to carried out upon TRIGA reactors, which aim to verify the real feasibility of the ADS (Accelerator Driven System) concept, it is essential to build a numerical tool able to simulate the dynamic behaviour of the reactor in subcritical configuration. This model developed to support the design of subcritical experiments and the safety analysis of the reactor, as a first step has to be assessed against the experimental data available for the critical reactor. To this purpose the thermal-hydraulic/ neutronic numerical model based on the RELAP5/PARCS coupled-code is been tested against the experimental reactivity transients conducted on the RC1-TRIGA reactor at the ENEA Casaccia Research Center in forecast of the TRADE (TRIGA Accelerator Driven Experiment) subcritical experience. The results of the calculations already performed show a qualitative good agreement with the experimental data and allow to address the future developments and improvements of the numerical model. (authors)

  12. State space modeling of reactor core in a pressurized water reactor

    SciTech Connect

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W.; Shamsuddin, Mustaffa; Abdullah, M. Adib

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  13. State space modeling of reactor core in a pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Ashaari, A.; Ahmad, T.; Shamsuddin, Mustaffa; M, Wan Munirah W.; Abdullah, M. Adib

    2014-07-01

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  14. TRIGA-SPEC: the prototype of MATS and LaSpec

    NASA Astrophysics Data System (ADS)

    Kaufmann, S.; Beyer, T.; Blaum, K.; Block, M.; Düllmann, Ch E.; Eberhardt, K.; Eibach, M.; Geppert, C.; Gorges, C.; Grund, J.; Hammen, M.; Krämer, J.; Nagy, Sz; Nörtershäuser, W.; Renisch, D.; Schneider, F.; Wendt, K.

    2015-04-01

    Investigation of short-lived nuclei is a challenging task that MATS and LaSpec will handle at the low energy branch of Super-FRS at FAIR. The groundwork for those experiments is laid-out already today at the TRIGA-SPEC facility as a powerful development platform located at the research reactor TRIGA Mainz. The latest status, new developments and first results of commissioning runs are presented here.

  15. Proposed modification of an instrumented TRIGA fuel element so that it may be handled with a standard TRIGA fuel handling tool

    SciTech Connect

    Doane, Harry J.

    1992-07-01

    Instrumented fuel elements whose thermocouples are no longer functional are still a useful source of reactor fuel. Their usefulness is hampered somewhat by the extension tubing that must extend above water level to keep the thermocouple extension leads dry and to keep pool water from interacting with the gas tight lead seal which is made below the lower coupling in the extension tubing. This facility proposes to modify an instrumented TRIGA fuel element by removing the extension tubing at the lower coupling and attaching to it a top end fixture that is normally supplied with a standard TRIGA fuel element. This would then allow movement of the modified fuel element with a standard TRIGA fuel handling tool. This paper will present the considerations involved in performing this modification and the presenter will solicit any useful information that might be contributed by attendees of the TRIGA Owners' Conference. (author)

  16. Reactor Dosimetry State of the Art 2008

    NASA Astrophysics Data System (ADS)

    Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

    2009-08-01

    Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

  17. Reactor Dosimetry State of the Art 2008

    NASA Astrophysics Data System (ADS)

    Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

    2009-08-01

    Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

  18. An Integrated Marine Propulsion System Utilising TRIGA{sup TM} Fuel

    SciTech Connect

    Manach, G.; Monnez, J-P.; Freeman, M.J.; Newell, A.; Brushwood, J.M.; Thompson, A.; Collins, C.; Scholes, N.; Hamilton, P.J.; Beeley, P.A.

    2004-07-01

    This paper describes the reactor physics, shielding, thermal hydraulics, reactor dynamics and safety studies conducted to develop a proposed Integrated Marine Propulsion System (IMPS) utilising TRIGA{sup TM} type uranium zirconium hydride fuel. The study has demonstrated that the IMPS plant is feasible and meets the design safety principles and safety criteria imposed on the study. (authors)

  19. High-Precision Mass Measurements At TRIGA-TRAP

    NASA Astrophysics Data System (ADS)

    Smorra, C.; Beyer, T.; Blaum, K.; Block, M.; Eberhardt, K.; Eibach, M.; Herfurth, F.; Ketelaer, J.; Knuth, K.; Nörtershäuser, W.; Nagy, Sz.

    2010-04-01

    In order to study neutron-rich nuclides far from the valley of stability as well as long-lived actinoids the double Penning-trap mass spectrometer TRIGA-TRAP has been recently installed at the research reactor TRIGA Mainz. Short-lived neutron-rich fission products are produced by thermal neutron-induced fission of an actinoid target installed close to the reactor core. A helium gas-jet system with carbon aerosol particles is used to extract the fission products to the experiment. The Penning trap system has already been commissioned. Off-line mass measurements are routinely performed using a recently developed laser ablation ion source, and the gas-jet system has been tested. An overview of the experiment and current status will be given.

  20. High-Precision Mass Measurements At TRIGA-TRAP

    SciTech Connect

    Smorra, C.; Eibach, M.; Beyer, T.; Blaum, K.; Block, M.; Herfurth, F.; Eberhardt, K.; Ketelaer, J.; Knuth, K.; Noertershaeuser, W.; Nagy, Sz.

    2010-04-30

    In order to study neutron-rich nuclides far from the valley of stability as well as long-lived actinoids the double Penning-trap mass spectrometer TRIGA-TRAP has been recently installed at the research reactor TRIGA Mainz. Short-lived neutron-rich fission products are produced by thermal neutron-induced fission of an actinoid target installed close to the reactor core. A helium gas-jet system with carbon aerosol particles is used to extract the fission products to the experiment. The Penning trap system has already been commissioned. Off-line mass measurements are routinely performed using a recently developed laser ablation ion source, and the gas-jet system has been tested. An overview of the experiment and current status will be given.

  1. Status of the TRIGA shipments to the INEEL from Asia

    SciTech Connect

    Tyacke, M.; George, W.; Petrasek, A.; Stump, R.C.; Patterson, J.

    1997-10-09

    This paper will report on preparations being made for returning Training, Research, Isotope, General Atomics (TRIGA) foreign research reactor (FRR) spent fuel from South Korea and Indonesia to the Idaho National Engineering and Environmental Laboratory (INEEL). The roles of US Department of Energy, INEEL, and NAC International in implementing a safe shipment are provided. Special preparations necessitated by making a shipment through a west coast port of the US to the INEEL will be explained. The institutional planning and actions needed to meet the unique political and operational environment for making a shipment from Asia to INEEL will be discussed. Facility preparation at both the INEEL and the FRRs is discussed. Cask analysis needed to properly characterize the various TRIGA configurations, compositions, and enrichments is discussed. Shipping preparations will include an explanation of the integrated team of spent fuel transportation specialists, and shipping resources needed to retrieve the fuel from foreign research reactor sites and deliver it to the INEEL.

  2. Experimental development of power reactor advanced controllers

    SciTech Connect

    Edwards, R.M.; Weng, C.K.; Lindsay, R.W.

    1992-06-01

    A systematic approach for developing and verifying advanced controllers with potential application to commercial nuclear power plants is suggested. The central idea is to experimentally demonstrate an advanced control concept first on an ultra safe research reactor followed by demonstration on a passively safe experimental power reactor and then finally adopt the technique for improving safety, performance, reliability and operability at commercial facilities. Prior to completing an experimental sequence, the benefits and utility of candidate advanced controllers should be established through theoretical development and simulation testing. The applicability of a robust optimal observer-based state feedback controller design process for improving reactor temperature response for a TRIGA research reactor, Liquid Metal-cooled Reactor (LMR), and a commercial Pressurized Water Reactor (PWR) is presented to illustrate the potential of the proposed experimental development concept.

  3. Experimental development of power reactor advanced controllers

    SciTech Connect

    Edwards, R.M. . Dept. of Nuclear Engineering); Weng, C.K. . Dept. of Mechanical Engineering); Lindsay, R.W. )

    1992-01-01

    A systematic approach for developing and verifying advanced controllers with potential application to commercial nuclear power plants is suggested. The central idea is to experimentally demonstrate an advanced control concept first on an ultra safe research reactor followed by demonstration on a passively safe experimental power reactor and then finally adopt the technique for improving safety, performance, reliability and operability at commercial facilities. Prior to completing an experimental sequence, the benefits and utility of candidate advanced controllers should be established through theoretical development and simulation testing. The applicability of a robust optimal observer-based state feedback controller design process for improving reactor temperature response for a TRIGA research reactor, Liquid Metal-cooled Reactor (LMR), and a commercial Pressurized Water Reactor (PWR) is presented to illustrate the potential of the proposed experimental development concept.

  4. Implementation of an aerodynamic lens for TRIGA-SPEC

    NASA Astrophysics Data System (ADS)

    Grund, J.; Düllmann, Ch. E.; Eberhardt, K.; Nagy, Sz.; van de Laar, J. J. W.; Renisch, D.; Schneider, F.

    2016-06-01

    We report on the optimization of the gas-jet system employed to couple the TRIGA-SPEC experiment to the research reactor TRIGA Mainz. CdI2 aerosol particles suspended in N2 as carrier gas are used for an effective transport of fission products from neutron induced 235 U fission from the target chamber to a surface ion source. Operating conditions of the gas-jet were modified to enable the implementation of an aerodynamic lens, fitting into the limited space available in front of the ion source. The lens boosts the gas-jet efficiency by a factor of 4-10. The characterization of the gas-jet system as well as the design of the aerodynamic lens and efficiency studies are presented and discussed.

  5. TRIGA FUEL PHASE I AND II CRITICALITY CALCULATION

    SciTech Connect

    L. Angers

    1999-11-23

    The purpose of this calculation is to characterize the criticality aspect of the codisposal of TRIGA (Training, Research, Isotopes, General Atomic) reactor spent nuclear fuel (SNF) with Savannah River Site (SRS) high-level waste (HLW). The TRIGA SNF is loaded into a Department of Energy (DOE) standardized SNF canister which is centrally positioned inside a five-canister defense SRS HLW waste package (WP). The objective of the calculation is to investigate the criticality issues for the WP containing the five SRS HLW and DOE SNF canisters in various stages of degradation. This calculation will support the analysis that will be performed to demonstrate the viability of the codisposal concept for the Monitored Geologic Repository (MGR).

  6. Support vector machines for nuclear reactor state estimation

    SciTech Connect

    Zavaljevski, N.; Gross, K. C.

    2000-02-14

    Validation of nuclear power reactor signals is often performed by comparing signal prototypes with the actual reactor signals. The signal prototypes are often computed based on empirical data. The implementation of an estimation algorithm which can make predictions on limited data is an important issue. A new machine learning algorithm called support vector machines (SVMS) recently developed by Vladimir Vapnik and his coworkers enables a high level of generalization with finite high-dimensional data. The improved generalization in comparison with standard methods like neural networks is due mainly to the following characteristics of the method. The input data space is transformed into a high-dimensional feature space using a kernel function, and the learning problem is formulated as a convex quadratic programming problem with a unique solution. In this paper the authors have applied the SVM method for data-based state estimation in nuclear power reactors. In particular, they implemented and tested kernels developed at Argonne National Laboratory for the Multivariate State Estimation Technique (MSET), a nonlinear, nonparametric estimation technique with a wide range of applications in nuclear reactors. The methodology has been applied to three data sets from experimental and commercial nuclear power reactor applications. The results are promising. The combination of MSET kernels with the SVM method has better noise reduction and generalization properties than the standard MSET algorithm.

  7. Evaluating Russian space nuclear reactor technology for United States applications

    SciTech Connect

    Polansky, G.F.; Schmidt, G.L.; Voss, S.S.; Reynolds, E.L.

    1994-08-01

    Space nuclear power and nuclear electric propulsion are considered important technologies for planetary exploration, as well as selected earth orbit applications. The Nuclear Electric Propulsion Space Test Program (NEPSTP) was intended to provide an early flight demonstration of these technologies at relatively low cost through extensive use of existing Russian technology. The key element of Russian technology employed in the program was the Topaz II reactor. Refocusing of the activities of the Ballistic Missile Defense Organization (BMDO), combined with budgetary pressures, forced the cancellation of the NEPSTP at the end of the 1993 fiscal year. The NEPSTP was faced with many unique flight qualification issues. In general, the launch of a spacecraft employing a nuclear reactor power system complicates many spacecraft qualification activities. However, the NEPSTP activities were further complicated because the reactor power system was a Russian design. Therefore, this program considered not only the unique flight qualification issues associated with space nuclear power, but also with differences between Russian and United States flight qualification procedures. This paper presents an overview of the NEPSTP. The program goals, the proposed mission, the spacecraft, and the Topaz II space nuclear power system are described. The subject of flight qualification is examined and the inherent difficulties of qualifying a space reactor are described. The differences between United States and Russian flight qualification procedures are explored. A plan is then described that was developed to determine an appropriate flight qualification program for the Topaz II reactor to support a possible NEPSTP launch.

  8. Temperature feedback of TRIGA MARK-II fuel

    NASA Astrophysics Data System (ADS)

    Usang, M. D.; Minhat, M. S.; Rabir, M. H.; M. Rawi M., Z.

    2016-01-01

    We study the amount of temperature feedback on reactivity for the three types of TRIGA fuel i.. ST8, ST12 and LEU fuel, are used in the TRIGA MARK II reactor in Malaysia Nuclear Agency. We employ WIMSD-5B for the calculation of kin f for a single TRIGA fuel surrounded by water. Typical calculations of TRIGA fuel reactivity are usually limited to ST8 fuel, but in this paper our investigation extends to ST12 and LEU fuel. We look at the kin f of our model at various fuel temperatures and calculate the amount reactivity removed. In one instance, the water temperature is kept at room temperature of 300K to simulate sudden reactivity increase from startup. In another instance, we simulate the sudden temperature increase during normal operation where the water temperature is approximately 320K while observing the kin f at various fuel temperatures. For accidents, two cases are simulated. The first case is for water temperature at 370K and the other is without any water. We observe that the higher Uranium content fuel such as the ST12 and LEU have much smaller contribution to the reactivity in comparison to the often studied ST8 fuel. In fact the negative reactivity coefficient for LEU fuel at high temperature in water is only slightly larger to the negative reactivity coefficient for ST8 fuel in void. The performance of ST8 fuel in terms of negative reactivity coefficient is cut almost by half when it is in void. These results are essential in the safety evaluation of the reactor and should be carefully considered when choices of fuel for core reconfiguration are made.

  9. Development of TRIGA-based experimental device for fiber optics in-core instrumentation testing for VHTRs

    SciTech Connect

    Johns, J. M.; Tsvetkov, P. V.

    2012-07-01

    Given the harsh environments of high temperature reactors, new in-core instrumentation has to be developed, since existing approaches may fail prematurely in VHTRs. The paper discusses ongoing efforts to support progress of suitable advanced in-core instrumentation technologies and develop an experimental approach for evaluation of their performance within VHTRs via emulation of VHTR in-core conditions in TRIGA reactors. Successful completion of the presented computational analysis concludes the first phase of the project. As demonstrated, it is proposed to use a high temperature furnace with fluence equivalency in operating TRIGA reactors. (authors)

  10. A combined gamma scanning and optical inspection system for spent TRIGA fuel

    SciTech Connect

    Boeck, H.; Allmer, G.

    1990-07-01

    A multipurpose lead container is used to investigate both the burn-up and the mechanical condition of standard TRIGA fuel elements. Especially in view of ageing reactor cores, this equipment is important to determine the further use of a specific fuel element and, therefore, saves operational costs.

  11. 11. biennial U.S. TRIGA users' conference. Papers and abstracts

    SciTech Connect

    1988-07-01

    The Conference was devoted to different aspects of TRIGA reactors design, operation and applications. The main topics concerned fuel elements, control rod drive system; modelling of corrosion damage and other chemical and material studies; neutron flux measurements and spectrum; irradiation devices; fuel element failures; neutron radiography etc.

  12. Intense steady state neutron source. The CNR reactor

    SciTech Connect

    Difilippo, F.C.; Moon, R.M.; Gambill, W.R.; Moon, R.M.; Primm, R.T. III; West, C.D.

    1986-01-01

    The Center for Neutron Research (CNR) has been proposed in response to the needs - neutron flux, spectrum, and experimental facilities - that have been identified through workshops, studies, and discussions by the neutron-scattering, isotope, and materials irradiation research communities. The CNR is a major new experimental facility consisting of a reactor-based steady state neutron source of unprecedented flux, together with extensive facilities and instruments for neutron scattering, isotope production, materials irradiation, and other areas of research.

  13. Tri-Gas Pressurization System Testing and Modeling for Cryogenic Applications

    NASA Technical Reports Server (NTRS)

    Taylor, B.; Polsgrove, R.; Stephens, J.; Hedayat, A.

    2014-01-01

    The use of Tri-gas in rocket propulsion systems is somewhat of a new technology. This paper defines Tri-gas as a mixture of gases composed largely of helium with a small percentage of a stoichiometric mixture of hydrogen and oxygen. When exposed to a catalyst the hydrogen and oxygen in the mixture combusts, significantly raising the temperature of the mixture. The increase in enthalpy resulting from the combustion process significantly decreases the required quantity of gas needed to pressurize the ullage of the vehicle propellant tanks. The objective of this effort was to better understand the operating characteristics of Tri-gas in a pressurization system with low temperature applications. In conjunction with ongoing programs at NASA Marshall Space Flight Center, an effort has been undertaken to evaluate the operating characteristics of Tri-gas through modeling and bench testing. Through improved understanding of the operating characteristics, the risk of using this new technology in a launch vehicle propulsion system was reduced. Bench testing of Tri-gas was a multistep process that targeted gas characteristics and performance aspects that pose a risk to application in a pressurization system. Pressurization systems are vital to propulsion system performance. Keeping a target ullage pressure in propulsions tanks is necessary to supply propellant at the conditions and flow rates required to maintain desired engine functionality. The first component of testing consisted of sampling Tri-gas sources that had been stagnant for various lengths of time in order to determine the rate at which stratification takes place. Second, a bench test was set up in which Tri-gas was sent through a catalyst bed. This test was designed to evaluate the performance characteristics of Tri-gas, under low temperature inlet temperatures, in a flight-like catalyst bed reactor. The third, most complex, test examined the performance characteristics of Tri-gas at low temperature temperatures

  14. Multiple steady states in coupled flow tank reactors

    NASA Astrophysics Data System (ADS)

    Hunt, Katharine L. C.; Kottalam, J.; Hatlee, Michael D.; Ross, John

    1992-05-01

    Coupling between continuous-flow, stirred tank reactors (CSTR's), each having multiple steady states, can produce new steady states with different concentrations of the chemical species in each of the coupled tanks. In this work, we identify a kinetic potential ψ that governs the deterministic time evolution of coupled tank reactors, when the reaction mechanism permits a single-variable description of the states of the individual tanks; examples include the iodate-arsenous acid reaction, a cubic model suggested by Noyes, and two quintic models. Stable steady states correspond to minima of ψ, and unstable steady states to maxima or saddle points; marginally stable states typically correspond to saddle-node points. We illustrate the variation in ψ due to changes in the rate constant for external material intake (k0) and for exchange between tanks (kx). For fixed k0 values, we analyze the changes in numbers and types of steady states as kx increases from zero. We show that steady states disappear by pairwise coalescence; we also show that new steady states may appear with increasing kx, when the reaction mechanism is sufficiently complex. For fixed initial conditions, the steady state ultimately reached in a mixing experiment may depend on the exchange rate constant as a function of time, kx(t) : Adiabatic mixing is obtained in the limit of slow changes in kx(t) and instantaneous mixing in the limit as kx(t)→∞ while t remains small. Analyses based on the potential ψ predict the outcome of mixing experiments for arbitrary kx(t). We show by explicit counterexamples that a prior theory developed by Noyes does not correctly predict the instability points or the transitions between steady states of coupled tanks, to be expected in mixing experiments. We further show that the outcome of such experiments is not connected to the relative stability of steady states in individual tank reactors. We find that coupling may effectively stabilize the tanks. We provide

  15. Hydrogen and Oxygen Gas Production in the UT TRIGA Reflector

    SciTech Connect

    D. S. O'Kelly

    2000-11-12

    In December 1999, The University of Texas at Austin (UT) reported an unusual condition associated with the annular graphite reflector surrounding the Nuclear Engineering Teaching Laboratory (NETL) TRIGA reactor. The aluminum container encapsulating the graphite showed signs of bulging or swelling. Further, during an investigation of this occurrence, bubbles were detected coming from a weld in the aluminum. The gas composition was approximately 2:1 hydrogen to oxygen. After safety review and equipment fabrication, the reflector was successfully vented and flooded. The ratio of the gases produced is unusual, and the gas production mechanism has not yet been explained.

  16. Status of reactor shielding research in the United States

    SciTech Connect

    Bartine, D.E.

    1983-01-01

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

  17. Facility modernization Annular Core Research Reactor

    SciTech Connect

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

    1990-07-01

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

  18. Design and Testing of a Boron Carbide Capsule for Spectral Tailoring in Mixed-Spectrum Reactors

    SciTech Connect

    Greenwood, Lawrence R.; Wittman, Richard S.; Pierson, Bruce D.; Metz, Lori A.; Payne, Rosara F.; Finn, Erin C.; Friese, Judah I.

    2012-03-01

    A boron carbide capsule has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State University. Irradiations were conducted in pulsed mode and in continuous operation for up to 4 hours. A cadmium cover was used to reduce thermal heating. The neutron spectrum calculated with MCNP was found to be in good agreement with reactor dosimetry measurements using the STAY'SL computer code. The neutron spectrum resembles that of a fast reactor. Design of a capsule using boron carbide enriched in {sup 10}B shows that it is possible to produce a neutron spectrum similar to {sup 235}U fission.

  19. Unconventional digital reactor control without conventional programming

    SciTech Connect

    Edwards, R.M.; Johns, R.M.; Kenney, S.J.

    1995-12-31

    Recent advances in simulation technology have resulted in the capability to design, test, and implement advanced control algorithms without the need for the labor-intensive effort of writing and debugging of computer programs. This technology has been adopted for a program of experimental development of power reactor control, which is jointly sponsored by the National Science Foundation and the Electric Power Research Institute. The experimental reactor control test bed utilizes the General Atomic Mark III TRIGA reactor at the Penn State Breazeale reactor facility. Control experiments are conducted within the movable experiment technical specifications of the TRIGA. A digital controller with an experimental control algorithm is interfaced to a secondary control rod (SCR). The new technology presented in this paper utilizes a UNIX network-compatible microprocessor-based controller operating under the Wind River Systems VxWorks real-time operating system. The controller interfaces with the Math-works MATLAB/SIMULINK development environment and Real-Time Innovations 8 monitoring software remotely operated on a SPARC workstation.

  20. TRIGA MARK-II source term

    NASA Astrophysics Data System (ADS)

    Usang, M. D.; Hamzah, N. S.; J. B., Abi M.; M. Z., M. Rawi; Abu, M. P.

    2014-02-01

    ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences of results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.

  1. TRIGA MARK-II source term

    SciTech Connect

    Usang, M. D. Hamzah, N. S. Abi, M. J. B. Rawi, M. Z. M. Rawi Abu, M. P.

    2014-02-12

    ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences of results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.

  2. TRIGA spent-fuel storage criticality analysis

    SciTech Connect

    Ravnik, M.; Glumac, B.

    1996-06-01

    A criticality safety analysis of a pool-type storage for spent TRIGA Mark II reactor fuel is presented. Two independent computer codes are applied: the MCNP Monte Carlo code and the WIMS lattice cell code. Two types of fuel elements are considered: standard fuel elements with 12 wt% uranium concentration and FLIP fuel elements. A parametric study of spent-fuel storage lattice pitch, fuel element burnup, and water density is presented. Normal conditions and postulated accident conditions are analyzed. A strong dependence of the multiplication factor on the distance between the fuel elements and on the effective water density is observed. A multiplication factor <1 may be expected for an infinite array of fuel rods at center-to-center distances >6.5 cm, regardless of the fuel element type and burnup. At shorter distances, the subcriticality can be ensured only by adding absorbers to the array of fuel rods even if the fuel rods were burned to {approximately}20% burnup. The results of both codes agree well for normal conditions. The results show that WIMS may be used as a complement to the Monte Carlo code in some parts of the criticality analysis.

  3. Expansion of a test bed for advanced reactor monitoring and control

    SciTech Connect

    Edwards, R.M.

    2000-07-01

    In previously completed work, the Penn State TRIGA reactor was established as a test bed for monitoring and control research for nuclear reactors. The essential component of this research reactor application is a means for an experiment to change reactor power through an experimental changeable reactivity device (ECRD). An ECRD is implemented as a TRIGA reactor moveable experiment where an aluminum tube containing an absorber material is positioned within the central thimble of the reactor by an experimental setup. The test bed capabilities are now being expanded to enhance research for monitoring, operations, and control under a US Department of Energy Nuclear Engineering Education and Research (NEER) grant initiated in 1999. Areas in which the capabilities of the test bed are being expanded are (a) experimental computer hardware and software upgrades, (b) additional ECRDs, (c) power-reactor thermal-hydraulic simulation fidelity in a hybrid reactor simulator (HRS) application, and (d) incorporation of a thermal-hydraulic testloop in the HRS paradigm. This summary describes progress in (a) and (b).

  4. 76 FR 21928 - Washington State University; Facility Operating License No. R-76; Washington State University...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-04-19

    ... Modified TRIGA Nuclear Radiation Center Reactor (NRCR); Environmental Assessment and Finding of No... TRIGA Nuclear Radiation Center Reactor (NRCR), located in the Dodgen Research Facility on Roundtop Drive... of quarterly exposure measurements at 12 locations adjacent to the Nuclear Radiation Center and at...

  5. Solid state laser driver for an ICF reactor

    SciTech Connect

    Krupke, W.F.

    1988-01-01

    A conceptual design is presented of the main power amplifier of a multi-beamline, multi-megawatt solid state ICF reactor driver. Simultaneous achievement of useful beam quality and high average power is achieved by a proper choice of amplifier geometry. An amplifier beamline consists of a sequence of face-pumped rectangular slab gain elements, oriented at the Brewster angle relative to the beamline axis, and cooled on their large faces by helium gas that is flowing subsonically. The infrared amplifier output radiation is shifted to an appropriately short wavelength (<500 nm) using nonlinear crystals that are also gas cooled. We project an overall driver efficiency >10% (including all flow cooling input power) when the amplifiers are pumped by efficient high-power AlGaAs semiconductor laser diode arrays. 11 refs., 3 figs., 7 tabs.

  6. Fuel behavior comparison for a research reactor

    NASA Astrophysics Data System (ADS)

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

    2006-06-01

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

  7. Activation analysis using Cornell TRIGA

    SciTech Connect

    Hossain, Tim Z.

    1994-07-01

    A major use of the Cornell TRIGA is for activation analysis. Over the years many varieties of samples have been analyzed from a number of fields of interest ranging from geology, archaeology and textiles. More recently the analysis has been extended to high technology materials for applications in optical and semiconductor devices. Trace analysis in high purity materials like Si wafers has been the focus in many instances, while in others analysis of major/minor components were the goals. These analysis has been done using the delayed mode. Results from recent measurements in semiconductors and other materials will be presented. In addition the near future capability of using prompt gamma activation analysis using the Cornell cold neutron beam will be discussed. (author)

  8. Ground State and Excited State H-Atom Temperatures in a Microwave Plasma Diamond Deposition Reactor

    NASA Astrophysics Data System (ADS)

    Gicquel, A.; Chenevier, M.; Breton, Y.; Petiau, M.; Booth, J. P.; Hassouni, K.

    1996-09-01

    Ground electronic state and excited state H-atom temperatures are measured in a microwave plasma diamond deposition reactor as a function of a low percentage of methane introduced in the feed gas and the averaged input microwave power density. Ground state H-atom temperatures (T_H) and temperature of the H-atom in the n=3 excited state (T_{Hα}) are obtained from the measurements respectively of the excitation profile by Two-photon Allowed transition Laser Induced Fluorescence (TALIF) and the Hα line broadening by Optical Emission Spectroscopy (OES). They are compared to gas temperatures calculated with a 1D diffusive non equilibrium H{2} plasma flow model and to ground electronic state rotational temperatures of molecular hydrogen measured previously by Coherent Anti-Stokes Raman Spectroscopy.

  9. Validation of the Serpent 2 code on TRIGA Mark II benchmark experiments.

    PubMed

    Ćalić, Dušan; Žerovnik, Gašper; Trkov, Andrej; Snoj, Luka

    2016-01-01

    The main aim of this paper is the development and validation of a 3D computational model of TRIGA research reactor using Serpent 2 code. The calculated parameters were compared to the experimental results and to calculations performed with the MCNP code. The results show that the calculated normalized reaction rates and flux distribution within the core are in good agreement with MCNP and experiment, while in the reflector the flux distribution differ up to 3% from the measurements. PMID:26516989

  10. The Ohio State University Reactor Sharing Program [August 15, 2000 - May 31, 2001

    SciTech Connect

    Myser, Richard D.

    2002-01-04

    During the period from August 15, 2000 through may 31, 2001 the Ohio State University (OSU) Research Reactor participated in the Reactor Sharing Program by providing services to nine colleges and universities and four secondary school organizations. A total of about 17 faculty and 170 students utilized their facilities. The staff of the OSU Research Reactor is generally involved in four types of experiments at the Nuclear Reactor Laboratory. Included are introductions to nuclear research, neutron activation analysis, material irradiation, and classes that measure various reactor parameters.

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

    PubMed

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

    2011-01-01

    In order to verify and validate the computational methods for neutron flux calculation in TRIGA research reactor calculations, a series of experiments has been performed. The neutron activation method was used to verify the calculated neutron flux distribution in the TRIGA reactor. Aluminium (99.9 wt%)-Gold (0.1 wt%) foils (disks of 5mm diameter and 0.2mm thick) were irradiated in 33 locations; 6 in the core and 27 in the carrousel facility in the reflector. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and experimental normalized reaction rates in the core are in very good agreement for both isotopes indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux and reaction rate distribution in the reactor core. In the reflector however, the accuracy of the epithermal and thermal neutron flux distribution and attenuation is lower, mainly due to lack of information about the material properties of the graphite reflector surrounding the core, but the differences between measurements and calculations are within 10%. Since our computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of research reactor utilization. PMID:20855215

  12. Neutron beam characterization at the Neutron Radiography Reactor (NRAD)

    SciTech Connect

    Imel, G.R.; Urbatsch, T.; Pruett, D.P.; Ross, J.R.

    1990-01-01

    The Neutron Radiography Reactor (NRAD) is a 250-kW TRIGA Reactor operated by Argonne National Laboratory and is located near Idaho Falls, Idaho. The reactor and its facilities regarding radiography are detailed in another paper at this conference; this paper summarizes neutron flux measurements and calculations that have been performed to better understand and potentially improve the neutronics characteristics of the reactor.

  13. K/sub infinity/-meter concept verified via subcritical-critical TRIGA experiments

    SciTech Connect

    Ocampo Mansilla, H.

    1983-01-01

    This work presents a technique for building a device to measure the k/sub infinity/ of a spent nuclear fuel assembly discharged from the core of a nuclear power plant. The device, called a k/sub infinity/-meter, consists of a cross-shaped subcritical assembly, two artificial neutron sources, and two separate neutron counting systems. The central position of the subcritical assembly is used to measure k/sub infinity/ of the spent fuel assembly. The initial subcritical assembly is calibrated to determine its k/sub eff/ and verify the assigned k/sub infinity/ of a selected fuel assembly placed in the central position. Count rates are taken with the fuel assembly of known k/sub infinity/'s placed in the central position and then repeated with a fuel assembly of unknown k/sub infinity/ placed in the central position. The count rate ratio of the unknown fuel assembly to the known fuel assembly is used to determine the k/sub infinity/ of the unknown fuel assembly. The k/sub infinity/ of the unknown fuel assembly is represented as a polynomial function of the count rate ratios. The coefficients of the polynomial equation are determined using the neutronic codes LEOPARD and EXTERMINATOR-II. The analytical approach has been validated by performing several subcritical/critical experiments, using the Penn State Breazeale TRIGA Reactor (PSBR), and comparing the experimental results with the calculations.

  14. 77 FR 5281 - State-of-the-Art Reactor Consequence Analyses Reports

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-02-02

    ... COMMISSION State-of-the-Art Reactor Consequence Analyses Reports AGENCY: Nuclear Regulatory Commission... Reactor Consequence Analyses (SOARCA) Report,'' for public comment. The purpose of Draft NUREG-1935 is to report a pilot study of best estimate analyses of the offsite radiological health consequences...

  15. Transport model based on three-dimensional cross-section generation for TRIGA core analysis

    NASA Astrophysics Data System (ADS)

    Kriangchaiporn, Nateekool

    This dissertation addresses the development of a reactor core physics model based on 3-D transport methodology utilizing 3-D multigroup fuel lattice cross-section generation and core calculation for PSBR. The proposed 3-D transport calculation scheme for reactor core simulations is based on the TORT code. The methodology includes development of algorithms for 2-D and 3-D cross-section generation. The fine- and broad-group structures for the TRIGA cross-section generation problems were developed based on the CPXSD (Contributon and Point-wise Cross-Section Driven) methodology that selects effective group structure. Along with the study of cross section generation, the parametric studies for SN calculations were performed to evaluate the impact of the spatial meshing, angular, and scattering order variables and to obtain the suitable values for cross-section collapsing of the TRIGA cell problem. The TRIGA core loading 2 is used to verify and validate the selected effective group structures. Finally, the 13 group structure was selected to use for core calculations. The results agree with continuous energy for eigenvalues and normalized pin power distribution. The Monte Carlo solutions are used as the references.

  16. Identification of leaking TRIGA fuel elements

    SciTech Connect

    Bennion, John S.; Crawford, Kevan C.; Gansauge, Todd C.; Sandquist, Gary M.

    1990-07-01

    The 100 kW TRIGA Mark I Nuclear Reactor at the University of Utah achieved initial criticality in October, 1975. Previously irradiated fuel consisting of stainless-steel- and aluminum-clad elements was acquired from the University of Arizona and the U.S. Army's Harry Diamond Laboratories in Adelphi, Maryland. Past core configurations have been comprised of both types of fuel with the aluminum-clad elements normally restricted to outer hexagonal rings of the core to provide a large safety margin between actual fuel temperature and limits set forth in the facility Technical Specifications. On October 20, 1987, trace cesium-137 contamination was discovered during routine analysis of the ion-exchange resin in the demineralizer circuit. The presence of Cs-137 indicated a possible clad defect resulting in the leakage of fission products. Reactor operations were allowed only to assist in identifying the source of the leakage. Pool water samples obtained following a two-hour operation at full power were spectroscopically analyzed and found to contain very small amounts of short-lived noble gases (e.g., Kr-85m, Kr-87, Kr-88, Xe-138) and their decay daughter products (e.g., Rb-88, Cs-138). Samples of the gaseous effluent from the facility collected in activated charcoal canisters showed no indication of fission product contamination. The small amount of activity released to the pool water suggested that a single defective element was responsible for the leakage. The instrumented fuel element and the aluminum-clad fuel were initially suspected as sources of the leakage. A simple scheme was devised to identify the defective element by exchanging four or five elements from the core with fuel in storage and then operating the reactor at 90 kW power for two hours. A pool water sample was then taken and analyzed to determine if the damaged element had been removed from the core. This process was repeated several times until all of the aluminum-clad fuel and several stainless

  17. High-temperature Chemical Compatibility of As-fabricated TRIGA Fuel and Type 304 Stainless Steel Cladding

    SciTech Connect

    Dennis D. Keiser, Jr.; Jan-Fong Jue; Eric Woolstenhulme; Kurt Terrani; Glenn A. Moore

    2012-09-01

    Chemical interaction between TRIGA fuel and Type-304 stainless steel cladding at relatively high temperatures is of interest from the point of view of understanding fuel behavior during different TRIGA reactor transient scenarios. Since TRIGA fuel comes into close contact with the cladding during irradiation, there is an opportunity for interdiffusion between the U in the fuel and the Fe in the cladding to form an interaction zone that contains U-Fe phases. Based on the equilibrium U-Fe phase diagram, a eutectic can develop at a composition between the U6Fe and UFe2 phases. This eutectic composition can become a liquid at around 725°C. From the standpoint of safe operation of TRIGA fuel, it is of interest to develop better understanding of how a phase with this composition may develop in irradiated TRIGA fuel at relatively high temperatures. One technique for investigating the development of a eutectic phase at the fuel/cladding interface is to perform out-of-pile diffusion-couple experiments at relatively high temperatures. This information is most relevant for lightly irradiated fuel that just starts to touch the cladding due to fuel swelling. Similar testing using fuel irradiated to different fission densities should be tested in a similar fashion to generate data more relevant to more heavily irradiated fuel. This report describes the results for TRIGA fuel/Type-304 stainless steel diffusion couples that were annealed for one hour at 730 and 800°C. Scanning electron microscopy with energy- and wavelength-dispersive spectroscopy was employed to characterize the fuel/cladding interface for each diffusion couple to look for evidence of any chemical interaction. Overall, negligible fuel/cladding interaction was observed for each diffusion couple.

  18. Mixed enrichment core design for the NC State University PULSTAR Reactor

    SciTech Connect

    Mayo, C.W.; Verghese, K.; Huo, Y.G.

    1997-12-01

    The North Carolina State University PULSTAR Reactor license was renewed for an additional 20 years of operation on April 30, 1997. The relicensing period added additional years to the facility operating time through the end of the second license period, increasing the excess reactivity needs as projected in 1988. In 1995, the Nuclear Reactor Program developed a strategic plan that addressed the future maintenance, development, and utilization of the facility. Goals resulting from this plan included increased academic utilization of the facility in accordance with its role as a university research facility, and increased industrial service use in accordance with the mission of a land grant university. The strategic plan was accepted, and it is the intent of the College of Engineering to operate the PULSTAR Reactor as a going concern through at least the end of the current license period. In order to reach the next relicensing review without prejudice due to low excess reactivity, it is desired to maintain sufficient excess reactivity so that, if relicensed again, the facility could continue to operate without affecting users until new fuel assistance was provided. During the NC State University license renewal, the operation of the PULSTAR Reactor at the State University of New York at Buffalo (SUNY Buffalo) was terminated. At that time, the SUNY Buffalo facility had about 240 unused PULSTAR Reactor fuel pins with 6% enrichment. The objective of the work reported here was to develop a mixed enrichment core design for the NC State University PULSTAR reactor which would: (1) demonstrate that 6% enriched SUNY buffalo fuel could be used in the NC State University PULSTAR Reactor within the existing technical specification safety limits for core physics parameters; (2) show that use of this fuel could permit operating the NC State University PULSTAR Reactor to 2017 with increased utilization; and (3) assure that the decision whether or not to relicense the facility would

  19. Status of development and licensing support for advanced liquid metal reactors in the United States

    SciTech Connect

    Pedersen, D.R. ); Gyorey, G. )

    1991-01-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment.

  20. Status of development and licensing support for advanced liquid metal reactors in the United States

    SciTech Connect

    Pedersen, D.R.; Gyorey, G.

    1991-12-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment.

  1. Solid state oxygen anion and electron mediating membrane and catalytic membrane reactors containing them

    DOEpatents

    Schwartz, Michael; White, James H.; Sammels, Anthony F.

    2000-01-01

    This invention relates to gas-impermeable, solid state materials fabricated into membranes for use in catalytic membrane reactors. This invention particularly relates to solid state oxygen anion- and electron-mediating membranes for use in catalytic membrane reactors for promoting partial or full oxidation of different chemical species, for decomposition of oxygen-containing species, and for separation of oxygen from other gases. Solid state materials for use in the membranes of this invention include mixed metal oxide compounds having the brownmillerite crystal structure.

  2. Solid state oxygen anion and electron mediating membrane and catalytic membrane reactors containing them

    DOEpatents

    Schwartz, Michael; White, James H.; Sammells, Anthony F.

    2005-09-27

    This invention relates to gas-impermeable, solid state materials fabricated into membranes for use in catalytic membrane reactors. This invention particularly relates to solid state oxygen anion- and electron-mediating membranes for use in catalytic membrane reactors for promoting partial or full oxidation of different chemical species, for decomposition of oxygen-containing species, and for separation of oxygen from other gases. Solid state materials for use in the membranes of this invention include mixed metal oxide compounds having the brownmillerite crystal structure.

  3. Steady-state tokamak reactor with non-divertor impurity control: STARFIRE

    SciTech Connect

    Baker, C.C.

    1980-01-01

    STARFIRE is a conceptual design study of a commercial tokamak fusion electric power plant. Particular emphasis has been placed on simplifying the reactor concept by developing design concepts to produce a steady-state tokamak with non-divertor impurity control and helium ash removal. The concepts of plasma current drive using lower hybrid rf waves and a limiter/vacuum system for reactor applications are described.

  4. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

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

  5. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

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

  6. An approach to model reactor core nodalization for deterministic safety analysis

    NASA Astrophysics Data System (ADS)

    Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

    2016-01-01

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  7. NFR TRIGA package design review report

    SciTech Connect

    Clements, M.D.

    1994-08-26

    The purpose of this document is to compile, present and document the formal design review of the NRF TRIGA packaging. The contents of this document include: the briefing meeting presentations, package description, design calculations, package review drawings, meeting minutes, action item lists, review comment records, final resolutions, and released drawings. This design review required more than two meeting to resolve comments. Therefore, there are three meeting minutes and two action item lists.

  8. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

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

  9. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

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

  10. Optimization of glycerol fed-batch fermentation in different reactor states: a variable kinetic parameter approach.

    PubMed

    Xie, Dongming; Liu, Dehua; Zhu, Haoli; Zhang, Jianan

    2002-05-01

    To optimize the fed-batch processes of glycerol fermentation in different reactor states, typical bioreactors including 500-mL shaking flask, 600-mL and 15-L airlift loop reactor, and 5-L stirred vessel were investigated. It was found that by reestimating the values of only two variable kinetic parameters associated with physical transport phenomena in a reactor, the macrokinetic model of glycerol fermentation proposed in previous work could describe well the batch processes in different reactor states. This variable kinetic parameter (VKP) approach was further applied to model-based optimization of discrete-pulse feed (DPF) strategies of both glucose and corn steep slurry for glycerol fed-batch fermentation. The experimental results showed that, compared with the feed strategies determined just by limited experimental optimization in previous work, the DPF strategies with VKPs adjusted could improve glycerol productivity at least by 27% in the scale-down and scale-up reactor states. The approach proposed appeared promising for further modeling and optimization of glycerol fermentation or the similar bioprocesses in larger scales. PMID:12049203

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

    SciTech Connect

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

    1996-12-31

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

  12. REACTOR

    DOEpatents

    Spitzer, L. Jr.

    1962-01-01

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

  13. Utilizing a Russian space nuclear reactor for a United States space mission: Systems integration issues

    SciTech Connect

    Reynolds, E.; Schaefer, E.; Polansky, G.; Lacy, J.; Bocharov, A.

    1993-09-30

    The Nuclear Electric Propulsion Space Test Program (NEPSTP) has developed a cooperative relationship with several institutes of the former Soviet Union to evaluate Russian space hardware on a US spacecraft One component is the Topaz II Nuclear Power System; a built and flight qualified nuclear reactor that has yet to be tested in space. The access to the Topaz II reactor provides the NEPSTP with a rare opportunity; to conduct an early flight demonstration of nuclear electric propulsion at a relatively low cost. This opportunity, however, is not without challenges. Topaz II was designed to be compatible with Russian spacecraft and launch vehicles. It was manufactured and flight qualified by Russian techniques and standards and conforms to safety requirements of the former Soviet Union, not the United States. As it is desired to make minimal modifications to the Topaz II, integrating the reactor system with a United States spacecraft and launch vehicle presents an engineering challenge. This paper documents the lessons teamed regarding the integration of reactor based spacecraft and also some insight about integrating Russian hardware. It examines the planned integration flow along with specific reactor requirements that affect the spacecraft integration including American-Russian space system compatibility.

  14. Monte Carlo Simulation of the TRIGA Mark II Benchmark Experiment with Burned Fuel

    SciTech Connect

    Jeraj, Robert; Zagar, Tomaz; Ravnik, Matjaz

    2002-03-15

    Monte Carlo calculations of a criticality experiment with burned fuel on the TRIGA Mark II research reactor are presented. The main objective was to incorporate burned fuel composition calculated with the WIMSD4 deterministic code into the MCNP4B Monte Carlo code and compare the calculated k{sub eff} with the measurements. The criticality experiment was performed in 1998 at the ''Jozef Stefan'' Institute TRIGA Mark II reactor in Ljubljana, Slovenia, with the same fuel elements and loading pattern as in the TRIGA criticality benchmark experiment with fresh fuel performed in 1991. The only difference was that in 1998, the fuel elements had on average burnup of {approx}3%, corresponding to 1.3-MWd energy produced in the core in the period between 1991 and 1998. The fuel element burnup accumulated during 1991-1998 was calculated with the TRIGLAV in-house-developed fuel management two-dimensional multigroup diffusion code. The burned fuel isotopic composition was calculated with the WIMSD4 code and compared to the ORIGEN2 calculations. Extensive comparison of burned fuel material composition was performed for both codes for burnups up to 20% burned {sup 235}U, and the differences were evaluated in terms of reactivity. The WIMSD4 and ORIGEN2 results agreed well for all isotopes important in reactivity calculations, giving increased confidence in the WIMSD4 calculation of the burned fuel material composition. The k{sub eff} calculated with the combined WIMSD4 and MCNP4B calculations showed good agreement with the experimental values. This shows that linking of WIMSD4 with MCNP4B for criticality calculations with burned fuel is feasible and gives reliable results.

  15. Proposed power upgrade of the Hot Fuel Examination Facility's neutron radiography reactor. [NRAD reactor

    SciTech Connect

    Pruett, D.P.; Richards, W.J.; Heidel, C.C.

    1984-01-01

    The Hot Fuel Examination Facility, HFEF, is one of several facilities located at the Argonne Site. HFEF comprises a large hot cell where both non-destructive and destructive examination of highly-irradiated reactor fuels are conducted in support of the LMFBR program. One of the non-destructive examination techniques utilized at HFEF is neutron radiography. Neutron radiography is provided by the NRAD reactor facility, which is located beneath the HFEF hot cell. The NRAD reactor is a TRIGA reactor and is operated at a steady state power level of 250 kW solely for neutron radiography and the development of radiography techniques. When the NRAD facility was designed and constructed, an operating power level of 250 kW was considered to be adequate for obtaining radiographs of the type of specimens envisaged at that time. A typical radiograph required approximately a twenty-minute exposure time. Specimens were typically single fuel rods placed in an aluminum tray. Since that time, however, several things have occurred that have tended to increase radiography exposure times to as much as 90 minutes each. In order to decrease exposure times, the reactor power level is to be increased from 250 kw to 1 MW. This increase in power will necessitate several engineering and design changes. These changes are described.

  16. TRIGA Mark II Criticality Benchmark Experiment with Burned Fuel

    SciTech Connect

    Persic, Andreja; Ravnik, Matjaz; Zagar, Tomaz

    2000-12-15

    The experimental results of criticality benchmark experiments performed at the Jozef Stefan Institute TRIGA Mark II reactor are presented. The experiments were performed with partly burned fuel in two compact and uniform core configurations in the same arrangements as were used in the fresh fuel criticality benchmark experiment performed in 1991. In the experiments, both core configurations contained only 12 wt% U-ZrH fuel with 20% enriched uranium. The first experimental core contained 43 fuel elements with average burnup of 1.22 MWd or 2.8% {sup 235}U burned. The last experimental core configuration was composed of 48 fuel elements with average burnup of 1.15 MWd or 2.6% {sup 235}U burned. The experimental determination of k{sub eff} for both core configurations, one subcritical and one critical, are presented. Burnup for all fuel elements was calculated in two-dimensional four-group diffusion approximation using the TRIGLAV code. The burnup of several fuel elements was measured also by the reactivity method.

  17. Study of the effect of {sup 135}Xe poison on the temperature coefficient of TRIGA fuel

    SciTech Connect

    Iorgulis, Constantin

    1992-07-01

    A study of the influence of {sup 135}Xe on the prompt negative temperature coefficient of the 14-MW Romanian TRIGA reactor has been performed. Because of its large absorption cross section below 0.1 eV, we expected that {sup 135}Xe might make a positive contribution to the temperature coefficient because the higher-energy neutrons are less likely to be absorbed by the Xe. This effect would be largest about 16 hours after reactor shutdown. In order to investigate this phenomenon, we have performed cell and core calculations for various fuel temperatures, burnups, and {sup 135}Xe levels. These calculations indeed show a positive contribution of {sup 135}Xe to the temperature coefficient, especially for high burnups, where little {sup 167}Er remains to absorb the higher-energy neutrons. Work is in progress to evaluate the effect of the smaller negative temperature coefficient on the consequences of reactivity insertion accidents in unfavorable situations of {sup 135}Xe poisoning of the Romanian TRIGA core. (author)

  18. Radionuclide mass inventory, activity, decay heat, and dose rate parametric data for TRIGA spent nuclear fuels

    SciTech Connect

    Sterbentz, J.W.

    1997-03-01

    Parametric burnup calculations are performed to estimate radionuclide isotopic mass and activity concentrations for four different Training, Research, and Isotope General Atomics (TRIGA) nuclear reactor fuel element types: (1) Aluminum-clad standard, (2) Stainless Steel-clad standard, (3) High-enrichment Fuel Life Improvement Program (FLIP), and (4) Low-enrichment Fuel Life Improvement Program (FLIP-LEU-1). Parametric activity data are tabulated for 145 important radionuclides that can be used to generate gamma-ray emission source terms or provide mass quantity estimates as a function of decay time. Fuel element decay heats and dose rates are also presented parametrically as a function of burnup and decay time. Dose rates are given at the fuel element midplane for contact, 3.0-feet, and 3.0-meter detector locations in air. The data herein are estimates based on specially derived Beginning-of-Life (BOL) neutron cross sections using geometrically-explicit TRIGA reactor core models. The calculated parametric data should represent good estimates relative to actual values, although no experimental data were available for direct comparison and validation. However, because the cross sections were not updated as a function of burnup, the actinide concentrations may deviate from the actual values at the higher burnups.

  19. COMSOL Simulations for Steady State Thermal Hydraulics Analyses of ORNL s High Flux Isotope Reactor

    SciTech Connect

    Khane, Vaibhav B; Jain, Prashant K; Freels, James D

    2012-01-01

    Simulation models for steady state thermal hydraulics analyses of Oak Ridge National Laboratory s High Flux Isotope Reactor (HFIR) have been developed using the COMSOL Multiphysics simulation software. A single fuel plate and coolant channel of each type of HFIR fuel element was modeled in three dimensions; coupling to adjacent plates and channels was accounted for by using periodic boundary conditions. The standard k- turbulence model was used in simulating turbulent flow with conjugate heat transfer. The COMSOL models were developed to be fully parameterized to allow assessing impacts of fuel fabrication tolerances and uncertainties related to low enriched uranium (LEU) fuel design and reactor operating parameters. Heat source input for the simulations was obtained from separate Monte Carlo N Particle calculations for the axially non-contoured LEU fuel designs at the beginning of the reactor cycle. Mesh refinement studies have been performed to calibrate the models against the pressure drop measured across the HFIR core.

  20. Design and testing of a 10B4C capsule for spectral-tailoring in mixed-spectrum reactors

    NASA Astrophysics Data System (ADS)

    Greenwood, L. R.; Wittman, R.; Metz, L. A.; Finn, E. C.; Friese, J. I.

    2014-04-01

    A boron carbide capsule highly enriched in 10B has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State University. New experiments show that enriching the boron to 96% B-10 results in additional absorption of neutrons in the resonance region thereby producing a neutron spectrum that is much closer to a pure 235U fission spectrum than measured previously with a natural boron carbide capsule. A cadmium outer cover was used to reduce thermal alpha heating. The neutron spectrum calculated with MCNP was found to be in very good agreement with measured activation rates from neutron fluence monitors.

  1. Design and Testing of a 10B4C Capsule for Spectral-Tailoring in Mixed-Spectrum Reactors

    SciTech Connect

    Greenwood, Lawrence R.; Wittman, Richard S.; Metz, Lori A.; Finn, Erin C.; Friese, Judah I.

    2014-04-11

    A boron carbide capsule highly enriched in 10B has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State University. New experiments show that enriching the boron to 96% B-10 results in additional absorption of neutrons in the resonance region thereby producing a neutron spectrum that is much closer to a pure 235U fission spectrum. A cadmium outer cover was used to reduce thermal heating. The neutron spectrum calculated with MCNP was found to be in very good agreement with measured activation rates from neutron fluence monitors.

  2. Participation in the United States Department of Energy Reactor Sharing Program

    SciTech Connect

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1992-12-01

    The University of Virginia Reactor Facility is an integral part of the Department of Mechanical, Aerospace and Nuclear Engineering and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and one of the most utilized university research reactor in the mid-Atlantic states. A major objective of this facility is to support educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor Sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1991 through August 1992. This report is also to final report under this contract. Previous annual reports should be consulted if any information from those periods is desired. Additional information about the programs conducted at UVA under this contract may be found in the yearly requests for additional funds which have been submitted.

  3. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

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

  4. PEBBLE: a two-dimensional steady-state pebble bed reactor thermal hydraulics code

    SciTech Connect

    Vondy, D.R.

    1981-09-01

    This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity. The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use.

  5. Assessment results of the South Korea TRIGA SNF to be shipped to INEEL

    SciTech Connect

    Cole, C.M.; Dirk, W.J.; Cottam, R.E.; Paik, S.T.

    1997-10-09

    This paper describes the Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) examination at the Seoul and the Taejon Research Reactor Facilities in South Korea. The examination was required before the SNF would be accepted for transportation and storage at the INEEL. The results of the aluminum and stainless steel clad TRIGA fuel examination have been summarized. A description of the examination team training, the examination work plan and examination equipment is also included. This paper also explains the technical basis for the examination and physical condition criteria used to determine what, if any, additional packaging would be required for transportation and for the receipt and storage of the fuel at the INEEL. This paper delineates the preparation activities prior to the fuel examinations and includes (1) collecting spent fuel data; (2) preparatory work by the Korean Atomic Energy Research Institute (KAERI) for fuel examination: (3) preparation of a radionuclide report, Radionuclide Mass Inventory, Activity, Decay Heat, and Dose Rate Parametric Data for TRIGA Spent Nuclear Fuels needed to provide input data for transportation and fuel acceptance at the Idaho National Engineering and Environmental Laboratory (INEEL); (4) gathering FRR Facility data; and (5) coordination between the INEEL and KAERI. Included, are the unanticipated conditions encountered in the unloading of fuel from the dry storage casks in Taejon in preparation for examination, a description of the damaged condition of the fuel removed from the casks, and the apparent cause of the damages. Lessons learned from all the activities are also addressed. A brief description of the preparatory work for the shipment of the spent fuel from Korea to INEEL is included.

  6. Computer modeling of the dynamic processes in the Maryland University Training Reactor - (MUTR)

    SciTech Connect

    White, Bernard H. IV; Ebert, David

    1988-07-01

    The simulator described in this paper models the behaviour of the Maryland University Training Reactor (MUTR). The reactor is a 250 kW, TRIGA reactor. The computer model is based on a system of five primary equations and eight auxiliary equations. The primary equations consist of the prompt jump approximation, a heat balance equation for the fuel and the moderator, and iodine and xenon buildup equations. For the comparison with the computer program, data from the reactor was acquired by using a personal computer (pc) which contained a Strawberry Tree data acquisition Card, connected to the reactor. The systems monitored by the pc were: two neutron detectors, fuel temperature, water temperature, three control rod positions and the period meter. The time differenced equations were programmed in the basic language. It has been shown by this paper, that the MUTR power rise from low power critical to high power, can be modelled by a relatively simple computer program. The program yields accurate agreement considering the simplicity of the program. The steady state error between the reactor and computer power is 4.4%. The difference in steady state temperatures, 112 deg. C and 117 deg. C, of the reactor and computer program, respectively, also yields a 4.5% error. Further fine tuning of the coefficients will yield higher accuracies.

  7. Design and testing of a boron carbide capsule for spectral-tailoring in mixed-spectrum reactors

    SciTech Connect

    Greenwood, L.R.; Wittman, R.; Pierson, B.P.; Metz, L.A.; Payne, R.; Finn, E.C.; Friese, J.I.

    2011-07-01

    A boron carbide capsule has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State Univ.. Irradiations were conducted in pulsed mode and in continuous operation for up to 4 h. A cadmium cover was used to reduce thermal heating. The neutron spectrum calculated with the Monte Carlo N-particle transport code was found to be in good agreement with reactor dosimetry measurements using the STAY'SL computer code. The neutron spectrum resembles that of a fast reactor. The design of a capsule using boron carbide fully enriched in {sup 10}B shows that it is possible to produce a neutron spectrum similar to that of {sup 235}U fission. (authors)

  8. The sharing of the Penn State Breazeale Reactor with other educational institutions

    SciTech Connect

    1995-12-31

    The Penn State Radiation Science and Engineering Center (RSEC) integrates the Breazeale Reactor and its affiliated laboratories and facilities on the University Park Campus. Penn State has the only nuclear reactor in Pennsylvania dedicated to research and education. Its faculty have pioneered industrial and research applications of radiation and radioisotopes. In addition, the center and its affiliated faculty have access to the multidisciplinary resources and expertise available within Penn State, one of the nation`s leading research universities. The goals of the Penn State Radiation Science and Engineering Center are to: incorporate radiation science and engineering services and facilities into a cohesive infrastructure; provide state-of-the-art academic instruction and laboratory experiences; provide facilities and assistance for academic research; provide technical, engineering, and other support services to RSEC users; generate new techniques, applications, and services for researchers in diverse disciplines; and serve the needs of academia and industry through RSEC services, faculty affiliations, and facilities that are not readily available elsewhere.

  9. Modeling and design of a new core-moderator assembly and neutron beam ports for the Penn State Breazeale Nuclear Reactor (PSBR)

    NASA Astrophysics Data System (ADS)

    Ucar, Dundar

    This study is for modeling and designing a new reactor core-moderator assembly and new neutron beam ports that aimed to expand utilization of a new beam hall of the Penn State Breazeale Reactor (PSBR). The PSBR is a part of the Radiation Science and Engineering Facility (RSEC) and is a TRIGA MARK III type research reactor with a movable core placed in a large pool and is capable to produce 1MW output. This reactor is a pool-type reactor with pulsing capability up to 2000 MW for 10-20 msec. There are seven beam ports currently installed to the reactor. The PSBR's existing core design limits the experimental capability of the facility, as only two of the seven available neutron beam ports are usable. The finalized design features an optimized result in light of the data obtained from neutronic and thermal-hydraulics analyses as well as geometrical constraints. A new core-moderator assembly was introduced to overcome the limitations of the existing PSBR design, specifically maximizing number of available neutron beam ports and mitigating the hydrogen gamma contamination of the neutron beam channeled in the beam ports. A crescent-shaped moderator is favored in the new PSBR design since it enables simultaneous use of five new neutron beam ports in the facility. Furthermore, the crescent shape sanctions a coupling of the core and moderator, which reduces the hydrogen gamma contamination significantly in the new beam ports. A coupled MURE and MCNP5 code optimization analysis was performed to calculate the optimum design parameters for the new PSBR. Thermal-hydraulics analysis of the new design was achieved using ANSYS Fluent CFD code. In the current form, the PSBR is cooled by natural convection of the pool water. The driving force for the natural circulation of the fluid is the heat generation within the fuel rods. The convective heat data was generated at the reactor's different operating powers by using TRIGSIMS, the fuel management code of the PSBR core. In the CFD

  10. Stress state dependence of in-reactor creep and swelling. Part 2: Experimental results

    NASA Astrophysics Data System (ADS)

    Hall, M. M., Jr.; Flinn, J. E.

    2010-01-01

    Irradiation creep constitutive equations, which were developed in Part I, are used here to analyze in-reactor creep and swelling data obtained ca. 1977-1979 as part of the US breeder reactor program. The equations were developed according to the principles of incremental continuum plasticity for the purpose of analyzing data obtained from a novel irradiation experiment that was conducted, in part, using Type 304 stainless steel that had been previously irradiated to significant levels of void swelling. Analyses of these data support an earlier observation that all stress states, whether tensile, compressive, shear or mixed, can affect both void swelling and interactions between irradiation creep and swelling. The data were obtained using a set of five unique multiaxial creep-test specimens that were designed and used for the first time in this study. The data analyses demonstrate that the constitutive equations derived in Part I provide an excellent phenomenological representation of the interactive creep and swelling phenomena. These equations provide nuclear power reactor designers and analysts with a first-of-its-kind structural analysis tool for evaluating irradiation damage-dependent distortion of complex structural components having gradients in neutron damage rate, temperature and stress state.

  11. The Intense Slow Positron Beam Facility at the NC State University PULSTAR Reactor

    SciTech Connect

    Hawari, Ayman I.; Moxom, Jeremy; Hathaway, Alfred G.; Brown, Benjamin; Xu, Jun

    2009-03-10

    An intense slow positron beam is in its early stages of operation at the 1-MW open-pool PULSTAR research reactor at North Carolina State University. The positron beam line is installed in a beam port that has a 30-cmx30-cm cross sectional view of the core. The positrons are created in a tungsten converter/moderator by pair-production using gamma rays produced in the reactor core and by neutron capture reactions in cadmium cladding surrounding the tungsten. Upon moderation, slow ({approx}3 eV) positrons that are emitted from the moderator are electrostatically extracted, focused and magnetically guided until they exit the reactor biological shield with 1-keV energy, approximately 3-cm beam diameter and an intensity exceeding 6x10{sup 8} positrons per second. A magnetic beam switch and transport system has been installed and tested that directs the beam into one of two spectrometers. The spectrometers are designed to implement state-of-the-art PALS and DBS techniques to perform positron and positronium annihilation studies of nanophases in matter.

  12. Research Reactor Benchmarks

    SciTech Connect

    Ravnik, Matjaz; Jeraj, Robert

    2003-09-15

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

  13. System and method for generating steady state confining current for a toroidal plasma fusion reactor

    DOEpatents

    Fisch, Nathaniel J.

    1981-01-01

    A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to establish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated in the plasma.

  14. System and method for generating steady state confining current for a toroidal plasma fusion reactor

    DOEpatents

    Bers, Abraham

    1981-01-01

    A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to estalish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated inthe plasma.

  15. Activity of synchronized cells of a steady-state biofilm recirculated reactor during xenobiotic biodegradation.

    PubMed Central

    Ascon-Cabrera, M A; Thomas, D; Lebeault, J M

    1995-01-01

    The maintenance of a steady-state biofilm in a continuous-flow fixed-bed reactor, as a consequence of the reproduction-detachment of cells (an interfacial cell physiology phenomenon of steady-state biofilm) during the biodegradation of 2,4,6-trichlorophenol by Pseudomonas cells, was determined. After cell adhesion on an open-pore glass support, the biofilm was formed in a packed-bed recirculated reactor. After the steady-state biofilm was reached, the mechanisms of the interfacial cell detachment (at the biofilm-liquid interface) were determined. It was established that (i) the hydrophobicity of immobilized sessile cells (parent cells) increased (from 50 to 80%) as the dilution rate increased, while the hydrophobicity of detached suspended cells (daughter cells) remained constant (about 45%); and (ii) the immediately detached suspended cells showed a synchronized growth in about three generations. These results indicate that (i) the immobilized sessile and suspended detached cells grew synchronically at the end and at the beginning of the cell cycle, respectively; and (ii) the hydrophobicity difference of immobilized sessile and suspended detached cells permitted the cells detachment. Therefore, it is probable that independent of shear stress (due to recirculated flow), the synchronized growth and hydrophobicity of cells (which vary during the cell cycle) are the main factors permitting the maintenance of a steady-state xenobiotic-degrading biofilm reactor (in which the overall accumulation of biofilm is determined by the average growth rate of the biofilm cells minus the rate of detachment of cells from the biofilm). PMID:7793923

  16. Steady state temperature profiles in two simulated liquid metal reactor fuel assemblies with identical design specifications

    SciTech Connect

    Levin, A.E.; Carbajo, J.J.; Lloyd, D.B.; Montgomery, B.H.; Rose, S.D.; Wantland, J.L.

    1985-01-01

    Temperature data from steady state tests in two parallel, simulated liquid metal reactor fuel assemblies with identical design specifications have been compared to determine the extent to which they agree. In general, good agreement was found in data at low flows and in bundle-center data at higher flows. Discrepancies in the data wre noted near the bundle edges at higher flows. An analysis of bundle thermal boundary conditions showed that the possible eccentric placement of one bundle within the housing could account for these discrepancies.

  17. Flux and instrumentation upgrade for the epithermal neutron beam facility at Washington State University.

    PubMed

    Nigg, D W; Venhuizen, J R; Wemple, C A; Tripard, G E; Sharp, S; Fox, K

    2004-11-01

    An epithermal neutron beam facility for preclinical neutron capture therapy research has been constructed at the Washington State University TRIGA research reactor installation. Subsequent to a recent upgrade, this new facility offers a high-purity epithermal beam with intensity on the order of 1.2x10(9)n/cm(2)s. Key features include a fluoride-based design for the neutron filtering and moderating components as well as a novel collimator design that allows ease of assembly and disassembly of the beamline components. PMID:15308181

  18. Flux and Instrumentation Upgrade for the Epithermal Neutron Beam Facility at Washington State University

    SciTech Connect

    David W. Nigg; J.R. Venhuizen; C.E. Wemple; G. E. Tripard; S. Sharp; K. Fox

    2004-11-01

    An epithermal neutron beam facility for preclinical neutron capture therapy research has been constructed at the Washington State University TRIGA research reactor installation. Subsequent to a recent upgrade, this new facility offers a high-purity epithermal beam with intensity on the order of 1.2×109 n/cm2 s. Key features include a fluoride-based design for the neutron filtering and moderating components as well as a novel collimator design that allows ease of assembly and disassembly of the beamline components.

  19. University Reactor Conversion Lessons Learned Workshop for Texas A&M University Nuclear Science Center Reactor

    SciTech Connect

    Eric C. Woolstenhulme; Dana M. Meyer

    2007-04-01

    The objectives of this meeting were to capture the observations, insights, issues, concerns, and ideas of those involved in the Texas A&M University Nuclear Science Center (TAMU NSC) TRIGA Reactor Conversion so that future efforts can be conducted with greater effectiveness, efficiency, and with fewer challenges. This workshop was held in conjunction with a similar workshop for the University of Florida Reactor Conversion. Some of the generic lessons from that workshop are included in this report for completeness.

  20. Irradiation facility at the TRIGA Mainz for treatment of liver metastases.

    PubMed

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

    2009-07-01

    The TRIGA Mark II reactor at the University of Mainz provides ideal conditions for duplicating BNCT treatment as performed in Pavia, Italy, in 2001 and 2003 [Pinelli, T., Zonta, A., Altieri, S., Barni, S., Braghieri, A., Pedroni, P., Bruschi, P., Chiari, P., Ferrari, C., Fossati, F., Nano, R., Ngnitejeu Tata, S., Prati, U., Ricevuti, G., Roveda, L., Zonta, C., 2002. TAOrMINA: from the first idea to the application to the human liver. In: Sauerwein et al. (Eds.), Research and Development in Neutron Capture Therapy. Proceedings of the 10th International Congress on Neutron Capture Therapy, Monduzzi editore, Bologna, pp. 1065-1072]. In order to determine the optimal parameters for the planned therapy and therefore for the design of the thermal column, calculations were conducted using the MCNP-code and the transport code ATTILA. The results of the parameter study as well as a possible configuration for the irradiation of the liver are presented. PMID:19394836

  1. An iterative approach for TRIGA fuel burn-up determination using nondestructive gamma-ray spectrometry.

    PubMed

    Wang, T K; Peir, J J

    2000-01-01

    The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products 97Zr/97Nb, 132I, and 140La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by reirradiating the spent fuels in a nuclear reactor. Based on the measured activities, 235U burn-up values can be deduced by iterative calculations. The complication caused by 239Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products 137Cs, 134Cs/137Cs ratio and 106Ru/137Cs ratio. PMID:10670930

  2. INVENTORY AND DESCRIPTION OF COMMERCIAL REACTOR FUELS WITHIN THE UNITED STATES

    SciTech Connect

    Vinson, D.

    2011-03-31

    There are currently 104 nuclear reactors in 31 states, operated by 51 different utilities. Operation of these reactors generates used fuel assemblies that require storage prior to final disposition. The regulatory framework within the United States (U.S.) allows for the licensing of used nuclear fuel storage facilities for an initial licensing period of up to 40 years with potential for license extensions in 40 years increments. Extended storage, for periods of up to 300 years, is being considered within the U.S. Therefore, there is an emerging need to develop the technical bases to support the licensing for long-term storage. In support of the Research and Development (R&D) activities required to support the technical bases, a comprehensive assessment of the current inventory of used nuclear fuel based upon publicly available resources has been completed that includes the most current projections of used fuel discharges from operating reactors. Negotiations with the nuclear power industry are ongoing concerning the willingness of individual utilities to provide information and material needed to complete the R&D activities required to develop the technical bases for used fuel storage for up to 300 years. This report includes a status of negotiations between DOE and industry in these regards. These negotiations are expected to result in a framework for cooperation between the Department and industry in which industry will provide and specific information on used fuel inventory and the Department will compensate industry for the material required for Research and Development and Testing and Evaluation Facility activities.

  3. Results from a scaled reactor cavity cooling system with water at steady state

    SciTech Connect

    Lisowski, D. D.; Albiston, S. M.; Tokuhiro, A.; Anderson, M. H.; Corradini, M. L.

    2012-07-01

    We present a summary of steady-state experiments performed with a scaled, water-cooled Reactor Cavity Cooling System (RCCS) at the Univ. of Wisconsin - Madison. The RCCS concept is used for passive decay heat removal in the Next Generation Nuclear Plant (NGNP) design and was based on open literature of the GA-MHTGR, HTR-10 and AVR reactor. The RCCS is a 1/4 scale model of the full scale prototype system, with a 7.6 m structure housing, a 5 m tall test section, and 1,200 liter water storage tank. Radiant heaters impose a heat flux onto a three riser tube test section, representing a 5 deg. radial sector of the actual 360 deg. RCCS design. The maximum heat flux and power levels are 25 kW/m{sup 2} and 42.5 kW, and can be configured for variable, axial, or radial power profiles to simulate prototypic conditions. Experimental results yielded measurements of local surface temperatures, internal water temperatures, volumetric flow rates, and pressure drop along the test section and into the water storage tank. The majority of the tests achieved a steady state condition while remaining single-phase. A selected number of experiments were allowed to reach saturation and subsequently two-phase flow. RELAP5 simulations with the experimental data have been refined during test facility development and separate effects validation of the experimental facility. This test series represents the completion of our steady-state testing, with future experiments investigating normal and off-normal accident scenarios with two-phase flow effects. The ultimate goal of the project is to combine experimental data from UW - Madison, UI, ANL, and Texas A and M, with system model simulations to ascertain the feasibility of the RCCS as a successful long-term heat removal system during accident scenarios for the NGNP. (authors)

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

    SciTech Connect

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

    2003-02-24

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

  5. Simultaneous carbon and nutrient removal in an airlift loop reactor under a limited filamentous bulking state.

    PubMed

    Jiang, Ming; Zhang, Yalei; Zhou, Xuefei; Su, Yimin; Zhang, Min; Zhang, Ke

    2013-02-01

    Airlift loop reactors (ALRs) are important bioreactors for wastewater treatment. However, few studies have investigated the application of an ALR for simultaneous carbon and nutrient removal, especially for activated sludge systems. This study evaluated the performance of integrated nitrogen, phosphorus and COD removal in an ALR with a low height-to-diameter ratio in a limited filamentous bulking (LFB) state (SVI of 180-220mL/g). The average removal efficiencies for COD, NH(4)(+)-N, TN and TP were 91%, 92%, 86% and 94%, respectively. Additional research showed that only under the LFB state, the appropriate distribution of dissolved oxygen inside the ALR was established to promote a well-balanced aerobic and anoxic/anaerobic state. In addition, the macro-gradient of the substrate concentration at the inlet and the heavier bio-P sludge density compensated for the proliferation of filaments. Hence, the stable LFB state was achieved by balancing the floc-forming bacteria and the filamentous bacteria in the ALR. PMID:23313686

  6. Design and Construction of the Ultracold Neutron Source at the NC State PULSTAR Research Reactor

    NASA Astrophysics Data System (ADS)

    Palmquist, Grant R.

    An ultracold neutron (UCN) source using solid deuterium is being constructed at the 1MWPULSTAR nuclear reactor on the campus of North Carolina State University. The final stages of assembly and commissioning are underway. The overall design, status of construction, and benchmarking measurements are presented. The UCN source design is based on detailed simulations including MCNP, UCN transport Monte Carlo, and computational fluid dynamics (CFD) simulation of the cryogenic systems. The source will be useful for developing UCN technologies, including guides and detectors, and in support of current projects including measurements of neutron beta-decay asymmetry coefficients and the electric dipolemoment of the neutron. The facility will also be available for testing new techniques using UCN in material and surface physics, as well as new fundamental physics measurements such as neutron lifetime and beta decay measurements. The expected experimental density of UCN/cm3 in a storage volume will be competitive with currently available sources, including those at significantly more powerful reactors.

  7. Solid-state track recorder neutron dosimetry in the Three-Mile Island Unit-2 reactor cavity

    NASA Astrophysics Data System (ADS)

    Gold, R.; Roberts, J. H.; Ruddy, F. H.; Preston, C. C.; McElroy, W. N.

    1985-04-01

    Solid state track recorder (SSTR) neutron dosimetry was conducted in the Three Mile Island Unit (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that there are at least 2 tonnes of fuels, which is roughly 4 fuel assemblies, at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity.

  8. ORIGEN2 calculations supporting TRIGA irradiated fuel data package

    SciTech Connect

    Schmittroth, F.A.

    1996-09-20

    ORIGEN2 calculations were performed for TRIGA spent fuel elements from the Hanford Neutron Radiography Facility. The calculations support storage and disposal and results include mass, activity,and decay heat. Comparisons with underwater dose-rate measurements were used to confirm and adjust the calculations.

  9. ESTABLISHING FINAL END STATE FOR A RETIRED NUCLEAR WEAPONS PRODUCTION REACTOR; COLLABORATION BETWEEN STAKEHOLDERS, REGULATORS, AND THE FEDERAL GOVERNMENT - 11052

    SciTech Connect

    Bergren, C.; Flora, M.; Belencan, H.

    2010-11-17

    The Savannah River Site (SRS) is a 310-square-mile United States Department of Energy nuclear facility located along the Savannah River (SRS) near Aiken, South Carolina. Nuclear weapons material production began in the early 1950s, utilizing five production reactors. In the early 1990s all SRS production reactor operations were terminated. The first reactor closure end state declaration was recently institutionalized in a Comprehensive Environmental Response and Compensation and Liability Act (CERCLA) Early Action Record of Decision. The decision for the final closure of the 318,000 square foot 105-P Reactor was determined to be in situ decommissioning (ISD). ISD is an acceptable and cost effective alternative to off-site disposal for the reactor building, which will allow for consolidation of remedial action wastes generated from other cleanup activities within the P Area. ISD is considered protective by the regulators, U. S. Environmental Protection Agency (US EPA) and the South Carolina Department of Health and Environmental Control (SCDHEC), public and stakeholders as waste materials are stabilized/immobilized, and radioactivity is allowed to naturally decay, thus preventing future exposure to the environment. Stakeholder buy-in was critical in the upfront planning in order to achieve this monumental final decision. Numerous public meetings and workshops were held in two different states (covering a 200 mile radius) with stakeholder and SRS Citizens Advisory Board participation. These meetings were conducted over an eight month period as the end state decision making progressed. Information provided to the public evolved from workshop to workshop as data became available and public input from the public meetings were gathered. ISD is being considered for the balance of the four SRS reactors and other hardened facilities such as the chemical Separation Facilities (canyons).

  10. ESTABLISHING FINAL END STATE FOR A RETIRED NUCLEAR WEAPONS PRODUCTION REACTOR; COLLABORATION BETWEEN STAKEHOLDERS, REGULATORS AND THE FEDERAL GOVERNMENT

    SciTech Connect

    Bergren, C

    2009-01-16

    The Savannah River Site (SRS) is a 310-square-mile United States Department of Energy nuclear facility located along the Savannah River (SRS) near Aiken, South Carolina. Nuclear weapons material production began in the early 1950s, utilizing five production reactors. In the early 1990s all SRS production reactor operations were terminated. The first reactor closure end state declaration was recently institutionalized in a Comprehensive Environmental Response and Compensation and Liability Act (CERCLA) Early Action Record of Decision. The decision for the final closure of the 318,000 square foot 105-P Reactor was determined to be in situ decommissioning (ISD). ISD is an acceptable and cost effective alternative to off-site disposal for the reactor building, which will allow for consolidation of remedial action wastes generated from other cleanup activities within the P Area. ISD is considered protective by the regulators, U. S. Environmental Protection Agency (US EPA) and the South Carolina Department of Health and Environmental Control (SCDHEC), public and stakeholders as waste materials are stabilized/immobilized, and radioactivity is allowed to naturally decay, thus preventing future exposure to the environment. Stakeholder buy-in was critical in the upfront planning in order to achieve this monumental final decision. Numerous public meetings and workshops were held in two different states (covering a 200 mile radius) with stakeholder and SRS Citizens Advisory Board participation. These meetings were conducted over an eight month period as the end state decision making progressed. Information provided to the public evolved from workshop to workshop as data became available and public input from the public meetings were gathered. ISD is being considered for the balance of the four SRS reactors and other hardened facilities such as the chemical processing canyons.

  11. Advances in reactor physics education: Visualization of reactor parameters

    SciTech Connect

    Snoj, L.; Kromar, M.; Zerovnik, G.

    2012-07-01

    Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

  12. University Reactor Sharing Program. Final report, September 30, 1992--September 29, 1994

    SciTech Connect

    Wehring, B.W.

    1995-01-01

    Over the past 20 years, the number of nuclear reactors on university campuses in the US declined from more than 70 to less than 40. Contrary to this trend, The University of Texas at Austin constructed a new reactor facility at a cost of $5.8 million. The new reactor facility houses a new TRIGA Mark II reactor which replaces an in-ground TRIGA Mark I reactor located in a 50-year old building. The new reactor facility was constructed to strengthen the instruction and research opportunities in nuclear science and engineering for both undergraduate and graduate students at The University of Texas. On January 17, 1992, The University of Texas at Austin received a license for operation of the new reactor. Initial criticality was achieved on March 12, 1992, and full power operation, on March 25, 1992. The UT-TRIGA research reactor provides hands-on education, multidisciplinary research and unique service activities for academic, medical, industrial, and government groups. Support by the University Reactor Sharing Programs increases the availability of The University of Texas reactor facility for use by other educational institutions which do not have nuclear reactors.

  13. State-of-the-art for liquid-level measurements applied to in-vessel coolant level for nuclear reactors

    SciTech Connect

    Anderson, R.L.

    1980-01-01

    The TMI-2 accident indicated that a direct indication of the liquid level in the reactor vessel would have told the operators that the core was being uncovered. This state-of-the-cost survey covered the following methods: heated thermocouple, differential pressure, ultrasonic, capacitance, microwave, time-domain reflectometry, and externally mounted radiation detectors. (DLC)

  14. 77 FR 27487 - License Amendment Request From The State University of New York, University of Buffalo Reactor...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-10

    ... accordance with the NRC E-Filing rule (72 FR 49139, August 28, 2007). The E-Filing process requires... COMMISSION License Amendment Request From The State University of New York, University of Buffalo Reactor....resource@nrc.gov . The University of Buffalo Decommissioning Plan and License Amendment Request...

  15. Impact of worm predation on pseudo-steady-state of the circulating fluidized bed biofilm reactor.

    PubMed

    Li, Ming; Nakhla, George; Zhu, Jesse

    2013-01-01

    This paper studies integrated simultaneous carbon and nitrogen removal as well as worm predation, in a circulating fluidized bed biofilm reactor (CFBBR) operated with an anoxic-aerobic bioparticle recirculation. A lab-scale CFBBR with a 8.5-liter reaction zone comprising 2L anoxic and 6.5L aerobic compartments was designed to evaluate the aquatic Oligochaete worm effect. Long-term (200 days) performance showed that stable and high-rate chemical oxygen demand (COD) with sodium acetate as the carbon source and total nitrogen (NH(4)Cl as nitrogen source) conversions were achieved simultaneously, with low sludge production of 0.082 g VSS (volatile suspended solids) g COD(-1) at pseudo-steady-state. Worm predation, which causes considerable sludge reduction of the bioparticle process, was studied. The results proved that the worm predation has a significant impact on the pseudo-steady-state performance of the CFBBR, decreasing biomass yield, decreasing oxygen concentration and increasing expanded bed height. PMID:23201510

  16. Tokamak burn cycle study: a data base for comparing long pulse and steady-state power reactors

    SciTech Connect

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K. Jr.; Hassanein, A.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1983-11-01

    Several distinct operating modes (conventional ohmic, noninductive steady state, internal transformer, etc.) have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics (current drive efficiency) and engineering (superior materials) which will help achieve these goals for different burn cycles.

  17. Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: A review of the state of the art

    SciTech Connect

    March-Leuba, J. ); Rey, J.M. )

    1992-01-01

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of unexpected'' instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities have been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a new and improved'' state of the art has emerged recently.

  18. Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: A review of the state of the art

    SciTech Connect

    March-Leuba, J.; Rey, J.M.

    1992-05-01

    This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of ``unexpected`` instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities have been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a ``new and improved`` state of the art has emerged recently.

  19. Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

    SciTech Connect

    Not Available

    1992-07-01

    Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

  20. Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

    SciTech Connect

    Licht, J. R.; Bergeron, A.; Dionne, B.; Van den Branden, G.; Kalcheva, S.; Sikik, E.; Koonen, E.

    2015-12-01

    BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.

  1. Design of a PGAA Facility at the TRIGA Mark III of ININ, Mexico

    SciTech Connect

    Rios-Martinez, C.; Paredes-Gutierrez, L.; Arias, E. Alemon; Ortiz-Romero, M.E.

    2001-06-17

    A thermal neutron prompt gamma activation analysis (PGAA) facility is being developed at the TRIGA-Mark III research reactor, located at the Nabor Carrillo Nuclear Center of the Mexican Institute for Nuclear Research. The PGAA facility is to be built at the exit of a 3.9-m-long radial beam port, which pierces the graphite core reflector. The measured thermal neutron flux at the beam port exit is 0.7 x 10{sup 8} n/cm{sup 2}s, with an epithermal neutron flux of 0.66 x 10{sup 8} n/cm{sup 2}s and a gamma-ray dose of 0.1 Sv/h at full reactor power. Under these circumstances, the extraction of a suitable thermal neutron beam becomes quite challenging. The neutron beam filtering and collimation systems are to be designed for a substantial reduction of the background source components in order to maximize the usable thermal neutron intensity. To obtain reasonable PGAA performance from a filtered low-intensity thermal neutron beam, a Compton suppression feature is added to the detection system. Representative suppressed and unsuppressed spectra of paraffin (hydrogen) neutron capture gamma rays are shown.

  2. Investigation of Coded Source Neutron Imaging at the North Carolina State University PULSTAR Reactor

    SciTech Connect

    Xiao, Ziyu; Mishra, Kaushal; Hawari, Ayman; Bingham, Philip R; Bilheux, Hassina Z; Tobin Jr, Kenneth William

    2010-10-01

    A neutron imaging facility is located on beam-tube #5 of the 1-MWth PULSTAR reactor at the North Carolina State University. An investigation has been initiated to explore the application of coded imaging techniques at the facility. Coded imaging uses a mosaic of pinholes to encode an aperture, thus generating an encoded image of the object at the detector. To reconstruct the image recorded by the detector, corresponding decoding patterns are used. The optimized design of coded masks is critical for the performance of this technique and will depend on the characteristics of the imaging beam. In this work, Monte Carlo (MCNP) simulations were utilized to explore the needed modifications to the PULSTAR thermal neutron beam to support coded imaging techniques. In addition, an assessment of coded mask design has been performed. The simulations indicated that a 12 inch single crystal sapphire filter is suited for such an application at the PULSTAR beam in terms of maximizing flux with good neutron-to-gamma ratio. Computational simulations demonstrate the feasibility of correlation reconstruction methods on neutron transmission imaging. A gadolinium aperture with thickness of 500 m was used to construct the mask using a 38 34 URA pattern. A test experiment using such a URA design has been conducted and the point spread function of the system has been measured.

  3. Burst wait time simulation of CALIBAN reactor at delayed super-critical state

    SciTech Connect

    Humbert, P.; Authier, N.; Richard, B.; Grivot, P.; Casoli, P.

    2012-07-01

    In the past, the super prompt critical wait time probability distribution was measured on CALIBAN fast burst reactor [4]. Afterwards, these experiments were simulated with a very good agreement by solving the non-extinction probability equation [5]. Recently, the burst wait time probability distribution has been measured at CEA-Valduc on CALIBAN at different delayed super-critical states [6]. However, in the delayed super-critical case the non-extinction probability does not give access to the wait time distribution. In this case it is necessary to compute the time dependent evolution of the full neutron count number probability distribution. In this paper we present the point model deterministic method used to calculate the probability distribution of the wait time before a prescribed count level taking into account prompt neutrons and delayed neutron precursors. This method is based on the solution of the time dependent adjoint Kolmogorov master equations for the number of detections using the generating function methodology [8,9,10] and inverse discrete Fourier transforms. The obtained results are then compared to the measurements and Monte-Carlo calculations based on the algorithm presented in [7]. (authors)

  4. Economic impacts on the United States of siting decisions for the international thermonuclear experimental reactor

    SciTech Connect

    Peerenboom, J.P.; Hanson, M.E.; Huddleston, J.R.

    1996-08-01

    This report presents the results of a study that examines and compares the probable short-term economic impacts of the International Thermonuclear Experimental Reactor (ITER) on the United States (U.S.) if (1) ITER were to be sited in the U.S., or (2) ITER were to be sited in one of the other countries that, along with the U.S., is currently participating in the ITER program. Life-cycle costs associated with ITER construction, operation, and decommissioning are analyzed to assess their economic impact. A number of possible U.S. host and U.S. non-host technology and cost-sharing arrangements with the other ITER Parties are examined, although cost-sharing arrangements and the process by which the Parties will select a host country and an ITER site remain open issues. Both national and local/regional economic impacts, as measured by gross domestic product, regional output, employment, net exports, and income, are considered. These impacts represent a portion of the complex, interrelated set of economic considerations that characterize U.S. host and U.S. non-host participation in ITER. A number of other potentially important economic and noneconomic considerations are discussed qualitatively.

  5. Participation in the United States Department of Energy Reactor Sharing Program. Final report, September 1980--August 1992

    SciTech Connect

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1992-12-01

    The University of Virginia Reactor Facility is an integral part of the Department of Mechanical, Aerospace and Nuclear Engineering and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and one of the most utilized university research reactor in the mid-Atlantic states. A major objective of this facility is to support educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor Sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1991 through August 1992. This report is also to final report under this contract. Previous annual reports should be consulted if any information from those periods is desired. Additional information about the programs conducted at UVA under this contract may be found in the yearly requests for additional funds which have been submitted.

  6. Participation in the United States Department of Energy University Reactor Instrumentation Program. Final report, September 1990--August 1993

    SciTech Connect

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1993-10-01

    The University of Virginia Reactor Facility is an integral part of the Department of Mechanical, Aerospace and Nuclear Engineering and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and one of the most utilized university research reactor in the mid-Atlantic states. A major objective of this facility is to support educational programs in the region. The University of Virginia has received support under the U.S. Department of Energy (DOE) University Reactor Instrumentation Program every year since 1990. The monies from this program have been used to purchase new equipment to replace outdated or inadequate safety-related instrumentation used in conjunction with reactor operations. This report documents the equipment purchased and the status of the installation and use of this equipment from September 1990 through August 1993. This report constitutes the final report for this project period.

  7. Changing concepts of geologic structure and the problem of siting nuclear reactors: examples from Washington State

    SciTech Connect

    Tabor, R.W.

    1986-09-01

    The conflict between regulation and healthy evolution of geological science has contributed to the difficulties of siting nuclear reactors. On the Columbia Plateau in Washington, but for conservative design of the Hanford reactor facility, the recognition of the little-understood Olympic-Wallowa lineament as a major, possibly still active structural alignment might have jeopardized the acceptability of the site for nuclear reactors. On the Olympic Peninsula, evolving concepts of compressive structures and their possible recent activity and the current recognition of a subducting Juan de Fuca plate and its potential for generating great earthquakes - both concepts little-considered during initial site selection - may delay final acceptance of the Satsop site. Conflicts of this sort are inevitable but can be accommodated if they are anticipated in the reactor-licensing process. More important, society should be increasing its store of geologic knowledge now, during the current recess in nuclear reactor siting.

  8. Thermal-hydraulic modeling of the Pennsylvania State University Breazeale Nuclear Reactor (PSBR)

    NASA Astrophysics Data System (ADS)

    Chang, Jong E.

    2005-11-01

    Earlier experiments determined that the Pennsylvania State University Breazeale Nuclear Reactor (PSBR) core is cooled, not by an axial flow, but rather by a strong cross flow due to the thermal expansion of the coolant. To further complicate the flow field, a nitrogen-16 (N-16) pump was installed above the PSBR core to mix the exiting core buoyant thermal plume in order to delay the rapid release of radioactive N-16 to the PSBR pool surface. Thus, the interaction between the N-16 jet flow and the buoyancy driven flow complicates the analysis of the flow distribution in the PSBR pool. The main objectives of this study is to model the thermal-hydraulic behavior of the PSBR core and pool. During this study four major things were performed including the Computational Fluid Dynamics (CFD) model for the PSBR pool, the stand-alone fuel rod model for a PSBR fuel rod, the velocity measurements in and around the PSBR core, and the temperature measurements in the PSBR pool. Once the flow field was predicted by the CFD model, the measurement devices were manufactured and calibrated based on the CFD results. The major contribution of this study is to understand and to explain the flow behavior in the PSBR subchannels and pool using the FLOW3D model. The stand-alone dynamic fuel rod model was developed to determine the temperature distribution inside a PSBR fuel rod. The stand-alone fuel rod model was coupled to the FLOW3D model and used to predict the temperature behavior during steady-state and pulsing. The heat transfer models in the stand-alone fuel rod code are used in order to overcome the disadvantage of the CFD code, which does not calculate the mechanical stress, the gap conductance, and the two phase heat transfer. (Abstract shortened by UMI.)

  9. TRIGA: Telecommunications Protocol Processing Subsystem Using Reconfigurable Interoperable Gate Arrays

    NASA Technical Reports Server (NTRS)

    Pang, Jackson; Pingree, Paula J.; Torgerson, J. Leigh

    2006-01-01

    We present the Telecommunications protocol processing subsystem using Reconfigurable Interoperable Gate Arrays (TRIGA), a novel approach that unifies fault tolerance, error correction coding and interplanetary communication protocol off-loading to implement CCSDS File Delivery Protocol and Datalink layers. The new reconfigurable architecture offers more than one order of magnitude throughput increase while reducing footprint requirements in memory, command and data handling processor utilization, communication system interconnects and power consumption.

  10. Design and Validation of Optimized Feedforward with Robust Feedback Control of a Nuclear Reactor

    SciTech Connect

    Shaffer, Roman; He Weidong; Edwards, Robert M.

    2004-08-15

    Design applications for robust feedback and optimized feedforward control, with confirming results from experiments conducted on the Pennsylvania State University TRIGA reactor, are presented. The combination of feedforward and feedback control techniques complement each other in that robust control offers guaranteed closed-loop stability in the presence of uncertainties, and optimized feedforward offers an approach to achieving performance that is sometimes limited by overly conservative robust feedback control. The design approach taken in this work combines these techniques by first designing robust feedback control. Alternative methods for specifying a low-order linear model and uncertainty specifications, while seeking as much performance as possible, are discussed and evaluated. To achieve desired performance characteristics, the optimized feedforward control is then computed by using the nominal nonlinear plant model that incorporates the robust feedback control.

  11. Frontier of Fusion Research: Path to the Steady State Fusion Reactor by Large Helical Device

    NASA Astrophysics Data System (ADS)

    Motojima, Osamu

    2006-12-01

    The ITER, the International Thermonuclear Experimental Reactor, which will be built in Cadarache in France, has finally started this year, 2006. Since the thermal energy produced by fusion reactions divided by the external heating power, i.e., the Q value, will be larger than 10, this is a big step of the fusion research for half a century trying to tame the nuclear fusion for the 6.5 Billion people on the Earth. The source of the Sun's power is lasting steadily and safely for 8 Billion years. As a potentially safe environmentally friendly and economically competitive energy source, fusion should provide a sustainable future energy supply for all mankind for ten thousands of years. At the frontier of fusion research important milestones are recently marked on a long road toward a true prototype fusion reactor. In its own merits, research into harnessing turbulent burning plasmas and thereby controlling fusion reaction, is one of the grand challenges of complex systems science. After a brief overview of a status of world fusion projects, a focus is given on fusion research at the National Institute for Fusion Science (NIFS) in Japan, which is playing a role of the Inter University Institute, the coordinating Center of Excellence for academic fusion research and by the Large Helical Device (LHD), the world's largest superconducting heliotron device, as a National Users' facility. The current status of LHD project is presented focusing on the experimental program and the recent achievements in basic parameters and in steady state operations. Since, its start in a year 1998, a remarkable progress has presently resulted in the temperature of 140 Million degree, the highest density of 500 Thousand Billion/cc with the internal density barrier (IDB) and the highest steady average beta of 4.5% in helical plasma devices and the largest total input energy of 1.6 GJ, in all magnetic confinement fusion devices. Finally, a perspective is given of the ITER Broad Approach program

  12. A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

    SciTech Connect

    Reed, Mark; Parker, Ronald R.; Forget, Benoit

    2012-06-19

    This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

  13. A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

    NASA Astrophysics Data System (ADS)

    Reed, Mark; Parker, Ronald R.; Forget, Benoit

    2012-06-01

    This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

  14. Nuclear Reactors. Revised.

    ERIC Educational Resources Information Center

    Hogerton, John F.

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

  15. NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE

    SciTech Connect

    John D. Bess

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately +/-1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  16. Evaluation of a Reactor On-Line Uncertainty Monitoring System

    SciTech Connect

    Edwards, Robert M.; He, Weidong

    2001-06-17

    Robust control designs were developed to better match experimental conditions available in the TRIGA reactor. A first-order weighting function is specified for each operating range in the robust design to limit the maximum tracking error. Inclusion of the performance-weighting function as the on-line filter is discussed as a possible on-line performance-monitoring method. The scheme to evaluate an on-line uncertainty monitoring system for a robust reactor controller is shown. TRIGA reactor experiments were conducted to evaluate on-line performance-monitoring techniques. It is concluded that the observed robust-control performance-monitoring characteristics can be incorporated in an on-line decision-making process to choose appropriate robust control selection and enforcement.

  17. Nuclear reactor pulse tracing using a CdZnTe electro-optic radiation detector

    NASA Astrophysics Data System (ADS)

    Nelson, Kyle A.; Geuther, Jeffrey A.; Neihart, James L.; Riedel, Todd A.; Rojeski, Ronald A.; Ugorowski, Philip B.; McGregor, Douglas S.

    2012-07-01

    CdZnTe has previously been shown to operate as an electro-optic radiation detector by utilizing the Pockels effect to measure steady-state nuclear reactor power levels. In the present work, the detector response to reactor power excursion experiments was investigated. Peak power levels during an excursion were predicted to be between 965 MW and 1009 MW using the Fuchs-Nordheim and Fuchs-Hansen models and confirmed with experimental data from the Kansas State University TRIGA Mark II nuclear reactor. The experimental arrangement of the Pockels cell detector includes collimated laser light passing through a transparent birefringent crystal, located between crossed polarizers, and focused upon a photodiode. The birefringent crystal, CdZnTe in this case, is placed in a neutron beam emanating from a nuclear reactor beam port. After obtaining the voltage-dependent Pockels characteristic response curve with a photodiode, neutron measurements were conducted from reactor pulses with the Pockels cell set at the 1/4 and 3/4 wave bias voltages. The detector responses to nuclear reactor pulses were recorded in real-time using data logging electronics, each showing a sharp increase in photodiode current for the 1/4 wave bias, and a sharp decrease in photodiode current for the 3/4 wave bias. The polarizers were readjusted to equal angles in which the maximum light transmission occurred at 0 V bias, thereby, inverting the detector response to reactor pulses. A high sample rate oscilloscope was also used to more accurately measure the FWHM of the pulse from the electro-optic detector, 64 ms, and is compared to the experimentally obtained FWHM of 16.0 ms obtained with the 10B-lined counter.

  18. Transportation infrastructure upgrades in the South: A compilation of state plans for construction near nuclear reactor sites

    SciTech Connect

    Not Available

    1992-03-01

    There are currently 27 nuclear reactor sites located in the southern region. In many instances, the most practicable modes of transportation of spent nuclear fuel from these sites we through the use of highway and rail systems. These two transportation modes have important differences that affect their applicability; chief among these, perhaps, is the fact that while highway systems are publicly owned and maintained rail lines are owned by private entities. For this reason, track condition and maintenance, usage rates and other aspects of rail transport can vary widely. This report reviews southern state, department plans for infrastructure upgrades in the vicinity of nuclear reactor sites. This report includes a summary of planned modifications to bridges, access highways, and rail spurs (where applicable) over the next five years. The information contained herein was gathered from interviews with officials within state departments of transportation. With few exceptions, the contact person was an official within the departmental planning division.

  19. Space reactor/organic Rankine conversion - A near-term state-of-the-art solution

    NASA Astrophysics Data System (ADS)

    Niggemann, R. E.; Lacey, D.

    The use of demonstrated reactor technology with organic Rankine cycle (ORC) power conversion can provide a low cost, minimal risk approach to reactor-powered electrical generation systems in the near term. Several reactor technologies, including zirconium hydride, EBR-II and LMFBR, have demonstrated long life and suitability for space application at the operating temperature required by an efficient ORC engine. While this approach would not replace the high temperature space reactor systems presently under development, it could be available in a nearer time frame at a low and predictable cost, allowing some missions requiring high power levels to be flown prior to the availability of advanced systems with lower specific mass. Although this system has relatively high efficiency, the heat rejection temperature is low, requiring a large radiator on the order of 3.4 sq m/kWe. Therefore, a deployable heat pipe radiator configuration will be required.

  20. Solid-state track recorder neutron dosimetry in the Three-Mile Island Unit-2 reactor cavity

    SciTech Connect

    Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.

    1985-04-01

    Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that there are at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity.

  1. Adaptive control of a packedbed reactor for the partial oxidation of n-butane to maleic anhydride; I. unsteady-state model and dynamics of the reactor

    SciTech Connect

    Not Available

    1984-01-01

    A nonadiabatic, fixed-bed catalytic reactor is analyzed on the basis of data for the highly exothermic partial oxidation of n-butane to maleic anhydride, in order to study multivariable adaptive control. The nonlinear partial differential equations describing the axial and radial gradients of concentration and temperature are converted into a set of nonlinear, ordinary differential and algebraic equations using orthogonal collocation, preserving the nonlinearity of the reaction term. These equations describe satisfactorily the steady state and dynamic behavior. This two-dimensional model enables the relationship between the time and temperature of reaction, and between the concentration of maleic anhydride and the hot-spot temperature to be expressed adequately by seconddegree, low-order transfer functions. This technique is applicable for any process of reaction in a packed bed.

  2. Program for the Analysis of Reactor Transients

    Energy Science and Technology Software Center (ESTSC)

    2002-01-29

    This program is designed for use in predicting the course of and consequence of nondestructive accidents in research and test reactor cores. It is intended primarily for the analysis of plate type research and test reactors and has been subjected to extensive comparisons with the SPERT I and SPERT II experiments. These comparisons were quite favorable for a wide range of transients up to and including melting of the clad. Favorable comparisons have also beenmore » made for TRIGA reactor pulses in pin geometry. The PARET/ANL code has been used by the RERTR (Reduced Enrichment Research and Test Reactor) Program for the safety evaluation of many of the candidate reactors for reduced enrichment.« less

  3. Program for the Analysis of Reactor Transients

    SciTech Connect

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

    2002-01-29

    This program is designed for use in predicting the course of and consequence of nondestructive accidents in research and test reactor cores. It is intended primarily for the analysis of plate type research and test reactors and has been subjected to extensive comparisons with the SPERT I and SPERT II experiments. These comparisons were quite favorable for a wide range of transients up to and including melting of the clad. Favorable comparisons have also been made for TRIGA reactor pulses in pin geometry. The PARET/ANL code has been used by the RERTR (Reduced Enrichment Research and Test Reactor) Program for the safety evaluation of many of the candidate reactors for reduced enrichment.

  4. 77 FR 26321 - Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-03

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

  5. Comparative study of pulsed and steady-state tokamak reactor burn cycles

    SciTech Connect

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1984-05-01

    Four distinct operating modes have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue in pulsed poloidal field coils; out-of-plant fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics and engineering which will help achieve these goals for different burn cycles.

  6. Recent improvements to steady-state thermal-hydraulic analysis of research reactors in the RERTR Program at ANL.

    SciTech Connect

    Olson, A. P.; Kalimullah; Feldman, E. E.; Nuclear Engineering Division

    2006-01-01

    Recent reactor conversion studies in the RERTR Program have required expansion or revision of modeling capabilities for steady state thermalhydraulic analysis. For example, some reactors operate in laminar flow, necessitating new correlations for Nusselt number and for friction loss. Others have single-sided heating of edge channels. And some have geometrical details that require new modeling approaches to either simulate or validate. Computational fluid dynamics was compared with the 2-dimensional approximation to heat flow used by the PLTEMP/ANL V3.0 code. A very systematic approach to hot channel factors is implemented. A closed-form solution is now used in flat-plate geometry to improve both speed and accuracy of the solution. Direct heating to clad and coolant is now included. The Groenveld table lookup method is now available for determination of CHF. Flow excursion prediction is updated. All of these improvements have been incorporated in the PLTEMP/ANL V3.0 code.

  7. TRIGA high wt -% LEU fuel development program. Final report

    SciTech Connect

    West, G.B.

    1980-07-01

    The principal purpose of this work was to investigate the characteristics of TRIGA fuel where the contained U-235 was in a relatively high weight percent (wt %) of LEU (low enriched uranium - enrichment of less than 20%) rather than a relatively low weight percent of HEU (high enriched uranium). Fuel with up to 45 wt % U was fabricated and found to be acceptable after metallurgical examinations, fission product retention tests and physical property examinations. Design and safety analysis studies also indicated acceptable prompt negative temperature coefficient and core lifetime characteristics for these fuels.

  8. Safety evaluation report related to the renewal of the operating license for the research reactor at North Carolina State University

    SciTech Connect

    1997-04-01

    This safety evaluation report (SER) summarizes the findings of a safety review conducted by the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation (NRR). The staff conducted this review in response to a timely application filed by North Carolina State University (the licensee or NCSU) for a 20-year renewal of Facility Operating License R-120 to continue to operate the NCSU PULSTAR research reactor. The facility is located in the Burlington Engineering Laboratory complex on the NCSU campus in Raleigh, North Carolina. In its safety review, the staff considered information submitted by the licensee (including past operating history recorded in the licensee`s annual reports to the NRC), as well as inspection reports prepared by NRC Region H personnel and first-hand observations. On the basis of this review, the staff concludes that NCSU can continue to operate the PULSTAR research reactor, in accordance with its application, without endangering the health and safety of the public. 16 refs., 31 figs., 7 tabs.

  9. Participation in the United States Department of Energy Reactor Sharing Program

    SciTech Connect

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1992-05-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics (to become the Department of Mechanical, Aerospace and Nuclear Engineering on July 1, 1992). As such, it is effectively used to support educational programs in engineering and science at the University of Virginia as well as those at other area colleges and universities. The expansion of support to educational programs in the mid-east region is a major objective. To assist in meeting this objective, the University of Virginia has been supported under the US Department of Energy (DOE) Reactor Sharing Program since 1978. Due to the success of the program, this proposal requests continued DOE support through August 1993.

  10. Design Feasible Area on Water Cooled Thorium Breeder Reactor in Equilibrium States

    SciTech Connect

    Sidik Permana; Naoyuki Takaki; Hiroshi Sekimoto

    2006-07-01

    Thorium as supplied fuel has good candidate for fuel material if it is converted into fissile material {sup 233}U which shows superior characteristics in the thermal region. The Shippingport reactor used {sup 233}U-Th fuel system, and the molten salt breeder reactor (MSBR) project showed that breeding is possible in a thermal spectrum. In the present study, feasibility of water cooled thorium breeder reactor is investigated. The key properties such as flux, {eta} value, criticality and breeding performances are evaluated for different moderator to fuel ratios (MFR) and burn-ups. The results show the feasibility of breeding for different MFR and burn-ups. The required {sup 233}U enrichment is about 2% - 9% as charge fuel. The lower MFR and the higher enrichment of {sup 233}U are preferable to improve the average burn-up; however the design feasible window is shrunk. This core shows the design feasible window especially in relation to MFR with negative void reactivity coefficient. (authors)

  11. Design Verification Report Neutron Radiography Facility (NRF) TRIGA Fuel Storage Systems

    SciTech Connect

    CARRELL, R.D.

    2002-01-31

    This report outlines the methods, procedures, and outputs developed during the Neutron Radiography Facility (NRF) Training, Research and Isotope Production, General Atomics (TRIGA) fuel storage system design and fabrication.

  12. Development of a steady state creep behavior model of polycrystalline tungsten for bimodal space reactor application

    SciTech Connect

    Purohit, A.; Hanan, N.A.; Bhattacharyya, S.K.; Gruber, E.E.

    1995-02-01

    The fuel element for one of the many reactor concepts being currently evaluated for bimodal applications in space consists of spherical fuel particles clad with tungsten or alloys of tungsten. The fuel itself consists of stabilized UO{sub 2}. One of the life limiting phenomena for the fuel element is failure of the cladding because of creep deformation. This report summarizes the information available in literature regarding the creep deformation of tungsten and its alloys and proposes a relation to be used for calculating the creep strains for elevated temperatures in the low stress region ({sigma} {le} 20 MPa). Also, results of the application of this creep relation to one of the reactor design concepts (NEBA-3) are discussed. Based on the traditional definition of creep deformation, the temperatures of 1500 K to 2900 K for tungsten and its alloys are considered to be in the {open_quotes}high{close_quotes} temperature range. In this temperature range, the rate controlling mechanisms for creep deformation are believed to be non-conservative motion of screw dislocations and short circuit diffusional paths. Extensive theoretical work on creep and in particular for creep of tungsten and its alloys have been reported in the literature. These theoretical efforts have produced complex mathematical models that require detailed materials properties. These relations, however, are not presently suitable for the creep analysis because of lack of consistent material properties required for their use. Variations in material chemistry and thermomechanical pre-treatment of tungsten have significant effects on creep and the mechanical properties. Analysis of the theoretical models and limited data indicates that the following empirical relation originally proposed by M. Jacox of INEL and the Air Force Phillips Laboratory, for calculating creep deformation of tungsten cladding, can be used for the downselection of preliminary bimodal reactor design concepts.

  13. A combined wet/dry sipping cell for investigating failed TRIGA fuel elements

    SciTech Connect

    Hammer, J.; Gallhammer, H.; Bock, H.

    1988-07-01

    Investigation for a failed TRIGA fuel element is performed with the help of a combined wet/dry sipping cell, which has been designed and fabricated at the Atominstitut Vienna. In this sipping cell a TRIGA fuel element can be studied for fission product release, both at normal and at elevated temperatures. This report describes the design features of the sipping cell and the fission product identification procedure with the help of a high purity Germanium detector and a multichannel analyzer.

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

    SciTech Connect

    Aden, V.G.; Kartashov, E.F.; Lukichev, V.A.

    1997-08-01

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

  15. Magnetic susceptibility as a direct measure of oxidation state in LiFePO4 batteries and cyclic water gas shift reactors.

    PubMed

    Kadyk, Thomas; Eikerling, Michael

    2015-08-14

    The possibility of correlating the magnetic susceptibility to the oxidation state of the porous active mass in a chemical or electrochemical reactor was analyzed. The magnetic permeability was calculated using a hierarchical model of the reactor. This model was applied to two practical examples: LiFePO4 batteries, in which the oxidation state corresponds with the state-of-charge, and cyclic water gas shift reactors, in which the oxidation state corresponds to the depletion of the catalyst. In LiFePO4 batteries phase separation of the lithiated and delithiated phases in the LiFePO4 particles in the positive electrode gives rise to a hysteresis effect, i.e. the magnetic permeability depends on the history of the electrode. During fast charge or discharge, non-uniform lithium distributionin the electrode decreases the hysteresis effect. However, the overall sensitivity of the magnetic response to the state-of-charge lies in the range of 0.03%, which makes practical measurement challenging. In cyclic water gas shift reactors, the sensitivity is 4 orders of magnitude higher and without phase separation, no hysteresis occurs. This shows that the method is suitable for such reactors, in which large changes of the magnetic permeability of the active material occurs. PMID:26156571

  16. Reactor performance and microbial community dynamics during solid-state anaerobic digestion of corn stover at mesophilic and thermophilic conditions.

    PubMed

    Shi, Jian; Wang, Zhongjiang; Stiverson, Jill A; Yu, Zhongtang; Li, Yebo

    2013-05-01

    Reactor performance and microbial community dynamics were investigated during solid state anaerobic digestion (SS-AD) of corn stover at mesophilic and thermophilic conditions. Thermophilic SS-AD led to faster and greater reductions of cellulose and hemicelluloses during the first 12 days compared to mesophilic SS-AD. However, accumulation of volatile fatty acids (VFAs) was 5-fold higher at thermophilic than mesophilic temperatures, resulting in a large pH drop during days 6-12 in the thermophilic reactors. Culture-based enumeration revealed 10-50 times greater populations of cellulolytic and xylanolytic microbes during thermophilic SS-AD than mesophilic SS-AD. DGGE analysis of PCR amplified 16S rRNA genes showed dynamic shifts, especially during the thermophilic SS-AD, of bacterial and archaeal communities over the 38 days of SS-AD as a result of acclimation of the initial seed microbial consortia to the lignocellulosic feedstock. The findings of this study can guide future studies to improve efficiency and stability of SS-AD. PMID:23567733

  17. Population dynamics in controlled unsteady-state systems: An application to the degradation of glyphosate in a sequencing batch reactor

    SciTech Connect

    Devarakonda, M.S.

    1988-01-01

    Control over population dynamics and organism selection in a biological waste treatment system provides an effective means of engineering process efficiency. Examples of applications of organism selection include control of filamentous organisms, biological nutrient removal, industrial waste treatment requiring the removal of specific substrates, and hazardous waste treatment. Inherently, full scale biological waste treatment systems are unsteady state systems due to the variations in the waste streams and mass flow rates of the substrates. Some systems, however, have the capacity to impose controlled selective pressures on the biological population by means of their operation. An example of such a system is the Sequencing Batch Reactor (SBR) which was the experimental system utilized in this research work. The concepts of organism selection were studied in detail for the biodegradation of a herbicide waste stream, with glyphosate as the target compound. The SBR provided a reactor configuration capable of exerting the necessary selective pressures to select and enrich for a glyphosate degrading population. Based on results for bench scale SBRs, a hypothesis was developed to explain population dynamics in glyphosate degrading systems.

  18. Comparison of the microbial communities in solid-state anaerobic digestion (SS-AD) reactors operated at mesophilic and thermophilic temperatures.

    PubMed

    Li, Yueh-Fen; Nelson, Michael C; Chen, Po-Hsu; Graf, Joerg; Li, Yebo; Yu, Zhongtang

    2015-01-01

    The microbiomes involved in liquid anaerobic digestion process have been investigated extensively, but the microbiomes underpinning solid-state anaerobic digestion (SS-AD) are poorly understood. In this study, microbiome composition and temporal succession in batch SS-AD reactors, operated at mesophilic or thermophilic temperatures, were investigated using Illumina sequencing of 16S rRNA gene amplicons. A greater microbial richness and evenness were found in the mesophilic than in the thermophilic SS-AD reactors. Firmicutes accounted for 60 and 82 % of the total Bacteria in the mesophilic and in the thermophilic SS-AD reactors, respectively. The genus Methanothermobacter dominated the Archaea in the thermophilic SS-AD reactors, while Methanoculleus predominated in the mesophilic SS-AD reactors. Interestingly, the data suggest syntrophic acetate oxidation coupled with hydrogenotrophic methanogenesis as an important pathway for biogas production during the thermophilic SS-AD. Canonical correspondence analysis (CCA) showed that temperature was the most influential factor in shaping the microbiomes in the SS-AD reactors. Thermotogae showed strong positive correlation with operation temperature, while Fibrobacteres, Lentisphaerae, Spirochaetes, and Tenericutes were positively correlated with daily biogas yield. This study provided new insight into the microbiome that drives SS-AD process, and the findings may help advance understanding of the microbiome in SS-AD reactors and the design and operation of SS-AD systems. PMID:25194839

  19. Criticality and characteristic neutronic analysis of a transient-state shockwave in a pulsed spherical gaseous uranium-hexafluoride reactor

    NASA Astrophysics Data System (ADS)

    Boles, Jeremiah Thomas

    The purpose of this study is to analyze the theoretical criticality of a spherical uranium-hexafluoride reactor with a transient, pulsed shockwave emanating from the center of the sphere in an outward-radial direction. This novel nuclear reactor design, based upon pulsed fission in a spherical enclosure is proposed for possible use in direct energy conversion, where the energy from fission products is captured through the use of electrostatic fields or through induction. An analysis of the dynamic behavior of the shockwave in this reactor is the subject of this thesis. As a shockwave travels through a fluid medium, the characteristics of the medium will change across the shockwave boundary. Pressure, temperature, and density are all affected by the shockwave. Changes in these parameters will affect the neutronic characteristics of a fissile medium. If the system is initially in a subcritical state, the increases in pressure, temperature, and density, all brought about by the introduction of the shockwave, will increase the reactivity of the nuclear system, creating a brief super critical state that will return to a subcritical state after the shockwave dissipates. Two major problems are required to be solved for this system. One is the effects of the shockwave on the gas, and the second is the resulting effects on system criticality. These problems are coupled due to the unique nature of the speed of the expanding shockwave in the uranium-hexafluoride medium and the energy imparted to the system by the shockwave with respect to the fissile uranium-hexafluoride. Using compressible flow and shockwave theories, this study determines the properties of the gaseous medium for reference points before, during, and behind the shockwave as it passes through the fissile medium. These properties include pressure changes, temperature changes, and density changes that occur to the system. Using the parameters calculated from the shockwave, the neutron transport equation is

  20. Description of TASHA: Thermal Analysis of Steady-State-Heat Transfer for the Advanced Neutron Source Reactor

    SciTech Connect

    Morris, D.G.; Chen, N.C.; Nelson, W.R.; Yoder, G.L.

    1996-10-01

    This document describes the code used to perform Thermal Analysis of Steady-State-Heat-Transfer for the Advanced Neutron Source (ANS) Reactor (TASHA). More specifically, the code is designed for thermal analysis of the fuel elements. The new code reflects changes to the High Flux Isotope Reactor steady-state thermal-hydraulics code. These changes were aimed at both improving the code`s predictive ability and allowing statistical thermal-hydraulic uncertainty analysis to be performed. A significant portion of the changes were aimed at improving the correlation package in the code. This involved incorporating more recent correlations for both single-phase flow and two-phase flow thermal limits, including the addition of correlations to predict the phenomenon of flow excursion. Since the code was to be used in the design of the ANS, changes were made to allow the code to predict limiting powers for a variety of thermal limits, including critical heat flux, flow excursion, incipient boiling, oxide spallation, maximum centerline temperature, and surface temperature equal to the saturation temperature. Statistical uncertainty analysis also required several changes to the code itself as well as changes to the code input format. This report describes these changes in enough detail to allow the reader to interpret code results and also to understand where the changes were made in the code programming. This report is not intended to be a stand alone report for running the code, however, and should be used in concert with the two previous reports published on the original code. Sample input and output files are also included to help accomplish these goals. In addition, a section is included that describes requirements for a new, more modem code that the project planned to develop.

  1. Compact Reactor

    SciTech Connect

    Williams, Pharis E.

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  2. Solid state proton and electron mediating membrane and use in catalytic membrane reactors

    DOEpatents

    White, James H.; Schwartz, Michael; Sammells, Anthony F.

    1998-01-01

    This invention provides catalytic proton and electron mediating membranes useful in catalytic reactors. The membranes have an oxidation and a reduction surface and comprise a single-phase mixed metal oxide material of the formula: AB.sub.1-x B'.sub.x O.sub.3-y wherein A is selected from Ca, Sr or Ba ions; B is selected from Ce, Tb, Pr, or Th ions; B' is selected from Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Al, Ga, or In ions, or combinations thereof; and x is greater than or equal to 0.02 and less than or equal to 0.5. The membranes can further comprise a catalyst on either the oxidation or reduction surface, or both. Membranes include those which are fabricated-by combining powders of metal oxides or metal carbonates of metal A ion, metal B ion and metal B' ion such that the stoichiometric ratio A:B:B' is 1:1-x:x where 0.2.ltoreq..times.0.5, repeatedly calcining and milling the combined powders until a single-phase material is obtained and pressing and sintering the singlephase material to obtain a membrane.

  3. Solid state proton and electron mediating membrane and use in catalytic membrane reactors

    DOEpatents

    White, J.H.; Schwartz, M.; Sammells, A.F.

    1998-10-13

    This invention provides catalytic proton and electron mediating membranes useful in catalytic reactors. The membranes have an oxidation and a reduction surface and comprise a single-phase mixed metal oxide material of the formula: AB{sub 1{minus}x}B{prime}{sub x}O{sub 3{minus}y} wherein A is selected from Ca, Sr or Ba ions; B is selected from Ce, Tb, Pr, or Th ions; B{prime} is selected from Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Al, Ga, or In ions, or combinations thereof; and x is greater than or equal to 0.02 and less than or equal to 0.5. The membranes can further comprise a catalyst on either the oxidation or reduction surface, or both. Membranes include those which are fabricated by combining powders of metal oxides or metal carbonates of metal A ion, metal B ion and metal B{prime} ion such that the stoichiometric ratio A:B:B{prime} is 1:1{minus}x:x where 0.2{<=}{times}0.5, repeatedly calcining and milling the combined powders until a single-phase material is obtained and pressing and sintering the single phase material to obtain a membrane. 6 figs.

  4. Membrane biofilm reactors for nitrogen removal: state-of-the-art and research needs.

    PubMed

    Hwang, Jong Hyuk; Cicek, Nazim; Oleszkiewicz, Jan A

    2009-01-01

    Historical developments up-to-date and operational challenges of membrane biofilm reactor (MBfR) were reviewed. A database of international, peer-reviewed journal articles regarding MBfR research from 1984 to 2008 was established and analyzed with a total of 107 papers. MBfR studies began to evolve in the early 1980s, since then the number of published papers increased steadily. After 2000, geographic locations where the research was conducted widened beyond North America and Europe to Asia. Research studies were divided into 4 categories and reviewed according to their main research focuses. In spite of the short history of MBfRs, studies have shown promising potential, possibly extending their application beyond nitrogen removal and organics removal. The MBfR research branched out to new fields including autotrophic denitrification. There are some important aspects of MBfRs that pose significant challenges to the application of this technology on a commercial scale in the near-future. The main challenge revolves around biofilm thickness and activity control. Further laboratory and demonstration scale studies on some of the proposed strategies for biofilm control are needed. Ultimately, more field studies with real wastewater should be performed to evaluate the resilience of the process in the face of flow and strength fluctuations, establishing optimum operational strategies. PMID:19934494

  5. High-intensity power-resolved radiation imaging of an operational nuclear reactor.

    PubMed

    Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

    2015-01-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

  6. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    NASA Astrophysics Data System (ADS)

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-10-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

  7. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    PubMed Central

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-01-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

  8. Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

    NASA Astrophysics Data System (ADS)

    Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

    2016-01-01

    The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel's center and surface, cladding, coolant temperatures as well as DNBR's values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

  9. Investigation of the low enrichment conversion of the Texas A and M Nuclear Science Center Reactor

    SciTech Connect

    Reuscher, J.A.

    1988-01-01

    The use of highly enriched uranium as a fuel research reactors is of concern due to the possibility of diversion for nuclear weapons applications. The Texas A M TRIGA reactor currently uses 70% enriched uranium in a FLIP (Fuel Life Improvement Program) fuel element manufactured by General Atomics. Thus fuel also contains 1.5 weight percent of erbium as a burnable poison to prolong useful core life. US university reactors that use highly enriched uranium will be required to covert to 20% or less enrichment to satisfy Nuclear Regulatory Commission requirements for the next core loading if the fuel is available. This investigation examined the feasibility of a material alternate to uranium-zirconium hydride for LEU conversion of a TRIGA reactor. This material is a beryllium oxide uranium dioxide based fuel. The theoretical aspects of core physics analyses were examined to assess the potential advantages of the alternative fuel. A basic model was developed for the existing core configuration since it is desired to use the present fuel element grid for the replacement core. The computing approach was calibrated to the present core and then applied to a core of BeO-UO{sub 2} fuel elements. Further calculations were performed for the General Atomics TRIGA low-enriched uranium zirconium hydride fuel.

  10. Dose Rate Calculations of Spent MTR-HEU Fuel Elements of the IAN-R1 Research Reactor

    NASA Astrophysics Data System (ADS)

    Sarta Fuentes, Jose Antonio

    2005-04-01

    With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safetly, a program was established at the Instituto de Ciencias Nucleares y Energìas Alternativas (INEA). This program included training, acquisition of hardware and sofware, design and construction of a decay pool, transfer of the spent HEU fuel elements into the decay pool and his final transport to Savanah River in United States. In this paper are presented external dose rates which were calculated for a standard spent MTR-HEU fuel element of the IAN-R1 Research Reactor. The calculations take in consideration the activity due to contributions of fission, activation and actinides products for each relevant radionuclide present in a standard spent MTR-HEU fuel. The datas obtained were the base for the respective dosimetric evaluations in the transfering operations of fuel elements into the decay pool and for shielding calculations in designing of the decay pool.

  11. Radionuclide Compositions and Total Activity of Spent MTR-HEU Fuel Elements of the IAN-R1 Research Reactor

    NASA Astrophysics Data System (ADS)

    Sarta, Josè A.; Castiblanco, Luis A.

    2005-05-01

    With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several calculations and tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safely a program was established at the Instituto de Ciencias Nucleares y Energìas Alternativas (INEA). This program included training, acquisition of hardware and software, design and construction of a decay pool, transfer of the spent HEU fuel elements into the decay pool and his final transport to Savannah River in United States. In this paper are presented data of activities calculated for each relevant radionuclide present in spent MTR-HEU fuel elements of the IAN-R1 Research Reactor and the total activity. The total activity calculated takes in consideration contributions of fission, activation and actinides products. The data obtained were the base for shielding calculations for the decay pool concerning the storage of spent MTR-HEU fuel elements and the respective dosimetric evaluations in the transferring operations of fuel elements into the decay pool.

  12. Radionuclide Compositions and Total Activity of Spent MTR-HEU Fuel Elements of the IAN-R1 Research Reactor

    SciTech Connect

    Sarta, Jose A.; Castiblanco, Luis A

    2005-05-24

    With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several calculations and tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safely a program was established at the Instituto de Ciencias Nucleares y Energias Alternativas (INEA). This program included training, acquisition of hardware and software, design and construction of a decay pool, transfer of the spent HEU fuel elements into the decay pool and his final transport to Savannah River in United States. In this paper are presented data of activities calculated for each relevant radionuclide present in spent MTR-HEU fuel elements of the IAN-R1 Research Reactor and the total activity. The total activity calculated takes in consideration contributions of fission, activation and actinides products. The data obtained were the base for shielding calculations for the decay pool concerning the storage of spent MTR-HEU fuel elements and the respective dosimetric evaluations in the transferring operations of fuel elements into the decay pool.

  13. State of the art and development trends in the production, properties, and uses of transplutonium elements at the Lenin Reactor Research Institute

    SciTech Connect

    Vasil'ev, V.Ya.; Gordeev, Ya.N.; Zinkovskii, V.I.; Karelin, E.A.; Klinov, A.V.; Kosulin, N.S.; Nikolaev, V.M.; Seleznev, A.G.; Timofeev, G.A.; Toporov, Yu.G.; Tsykanov, V.A.

    1987-03-01

    Some results are reported from the Reactor Research Institute on the properties, production methods, and uses of the transplutonium elements (TPE), which have been obtained between 1976 and 1982. A comprehensive technology has been devised for processing irradiated reactor targets, which provides high yields. Calculations and experiments have defined the optimum design of the high-frequency trap on the SM-2 reactor. Neutron sources have been made for general industrial and medical purposes containing californium 252. It is shown that curium occurs in the hexavalent state in aqueous solution. Measurements have been made on the resonant-level parameters of berkelium 249 and californium 249. TPE compounds with platinum-group metals have been made and examined. The properties of the TPE metals have also been examined. Future problems have been discussed.

  14. Dihydrogen gas emission of a 250kW(th) research reactor.

    PubMed

    Steinhauser, Georg; Villa, Mario

    2011-11-01

    The Vienna TRIGA pool-type reactor emits tiny gas bubbles at 250kW. They add up to a total volume of ∼2.4L in 7h of daily operation. The bubbles consist of nitrogen (72.5vol%), hydrogen (17.2%), oxygen plus argon (12.0%) and carbon dioxide (0.23%). The emission of constituents of air is caused by degassing of dissolved air in the hot regions of the reactor. Hydrogen results from neutron-induced radiolysis of the cooling water. This emission should be kept in mind for reasons of fire protection even for low-power reactors. PMID:21601464

  15. Application of an isothermal, three-phase catalytic reactor model to predict unsteady-state fixed-bed performance.

    PubMed

    Yang, Ji; Hand, David W; Hokanson, David R; Crittenden, John C

    2003-01-15

    CatReac, a three-phase catalytic mathematical model, was developed for analysis and optimization of the volatile reactor assembly used in International Space Station water processor. This wet oxidation process is used to remove low molecular weight contaminants such as acetic acid, acetone, ethanol, 1-propanol, 2-propanol, and propionic acid, which are not removed by the other treatment processes. The Langmuir-Hinshelwood (Hinshelwood, C. N. The Kinetics of Chemical Change in Gaseous Systems, 3rd ed.; Oxford: London, 1933; pp 301-347) isothermal adsorption expression was successfully used to describe the reaction kinetics of compounds on the catalyst surface for the compounds mentioned above. Small-column experiments combined with the use of the Arrhenius equation were successfully used to predict the Langmuir-Hinshelwood parameters under different temperatures for a temperature range from 93 to 149 degrees C. Full-scale and small-column experiments were successfully used to validate the model predictions for unsteady-state fixed-bed operations. PMID:12564919

  16. Recycled paper mill effluent treatment in a modified anaerobic baffled reactor: start-up and steady-state performance.

    PubMed

    Hassana, Siti Roshayu; Zwaina, Haider M; Zamana, Nastaein Qamaruz; Dahlanb, Irvan

    2014-01-01

    Start-up period is considered to be the most unstable and difficult stage in anaerobic process and usually takes a long time due to slow-degree adaptation of anaerobic microorganisms. In order to achieve a shorter start-up period, a novel modified anaerobic baffled reactor (MABR) has been developed in this study, where each modified baffle has its own characteristics (form/shape) to facilitate a treatment ofrecycled paper mill effluent (RPME). The results ofphysico-chemical characteristics showed that effluent from recycled paper mill consisted of 4328mgL-1 chemical oxygen demand (COD), 669mg L-1 biochemical oxygen demand and 501mg L-1 volatile fatty acid. It also consisted of variety of heavy metals such as zinc, magnesium, iron and nickel at concentrations of 1.39, 12.19, 2.39 and 0.72 mgL-1, respectively. Performance of MABR during the start-up period showed that methane production reached 34.7% with COD removal of 85% at steady state. The result indicates that MABR was successfully operated during the start-up period in treating RPME within a period of less than 30 days. PMID:24600868

  17. Environmental assessment for the manufacture and shipment of nuclear reactor fuel from the United States to Canada

    SciTech Connect

    Rangel, R.C.

    1999-02-01

    The US Department of Energy (DOE) has declared 41.9 tons (38 metric tons) of weapons-usable plutonium surplus to the United States` defense needs. A DOE Programmatic Environmental Impact Statement analyzed strategies for plutonium storage and dispositioning. In one alternative, plutonium as a mixed oxide (MOX) fuel would be irradiated (burned) in a reengineered heavy-water-moderated reactor, such as the Canadian CANDU design. In an Environmental Assessment (EA), DOE proposes to fabricate and transport to Canada a limited amount of MOX fuel as part of the Parallex (parallel experiment) Project. MOX fuel from the US and Russia would be used by Canada to conduct performance tests at Chalk River Laboratories. MOX fuel would be fabricated at Los Alamos National Laboratory and transported in approved container(s) to a Canadian port(s) of entry on one to three approved routes. The EA analyzes the environmental and human health effects from MOX fuel fabrication and transportation. Under the Proposed Action, MOX fuel fabrication would not result in adverse effects to the involved workers or public. Analysis showed that the shipment(s) of MOX fuel would not adversely affect the public, truck crew, and environment along the transportation routes.

  18. Computational and experimental physics performance characterization of the neutron capture therapy research facility at Washington State Univ

    SciTech Connect

    Nigg, D. W.; Sloan, P. E.; Venhuizen, J. R.; Wemple, C. A.; Tripard, G. E.; Fox, K.; Corwin, E.

    2006-07-01

    This paper summarizes the results of the final beam characterization measurements for a dual mode epithermal-thermal beam facility for neutron capture therapy research that was recently constructed at the Washington State Univ. TRIGA{sup TM} research reactor. The results show that the performance of the beam facility is consistent with the design computations and with international standards for the intended application. A useful epithermal neutron flux of 1.3 x 10{sup 9} n/cm{sup 2}-s is produced at the irradiation point with the beam in epithermal mode and shaped by a 10-cm circular aperture plate. When the beam is thermalized with approximately 34 cm of heavy water, the useful thermal flux at the irradiation point is approximately 3.5 x 10{sup 8} n/cm{sup 2}-s. The new WSU facility is one of only two such installations currently operating in the US. (authors)

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

    SciTech Connect

    CHOI, Kyung-Woo

    2008-01-15

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

  20. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    NASA Astrophysics Data System (ADS)

    Kulesza, Joel A.; Arzu Alpan, F.

    2016-02-01

    This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  1. Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

    SciTech Connect

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; Martinez, Patrick; Keller, Russ; Stanley, Floyd; Spencer, Khalil; Thomas, Mariam; Xu, Ning; Schappert, Michael; Fulwyler, James

    2015-10-26

    This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

  2. Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

    DOE PAGESBeta

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; Martinez, Patrick; Keller, Russ; Stanley, Floyd; Spencer, Khalil; Thomas, Mariam; Xu, Ning; Schappert, Michael; et al

    2015-10-26

    This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

  3. Effect of small-scale biomass gasification at the state of refractory lining the fixed bed reactor

    NASA Astrophysics Data System (ADS)

    Janša, Jan; Peer, Vaclav; Pavloková, Petra

    2016-06-01

    The article deals with the influence of biomass gasification on the condition of the refractory lining of a fixed bed reactor. The refractory lining of the gasifier is one part of the device, which significantly affects the operational reliability and durability. After removing the refractory lining of the gasifier from the experimental reactor, there was done an assessment how gasification of different kinds of biomass reflected on its condition in terms of the main factors affecting its life. Gasification of biomass is reflected on the lining, especially through sticking at the bottom of the reactor. Measures for prolonging the life of lining consist in the reduction of temperature in the reactor, in this case, in order to avoid ash fusion biomass which it is difficult for this type of gasifier.

  4. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Pressurized Water Reactor Standard Core Loading Benchmark Problem

    NASA Astrophysics Data System (ADS)

    Arzu Alpan, F.; Kulesza, Joel A.

    2016-02-01

    This paper compares contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a pressurized water reactor calculational benchmark problem with a standard out-in core loading. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission and used the Oak Ridge National Laboratory two-dimensional discrete ordinates code DORT and the BUGLE-93 cross-section library for the calculations. In this paper, a Westinghouse three-dimensional discrete ordinates code with parallel processing, the RAPTOR-M3G code was used. A variety of cross section libraries were used with RAPTOR-M3G including the BUGLE-93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory, and the broad-group ALPAN-VII.0 cross-section library developed at Westinghouse. In comparing the calculation-to-calculation reaction rates using the BUGLE-93 cross-section library at the thermal shield, pressure vessel, and cavity capsules, for eleven dosimetry reaction rates, a maximum relative difference of 5% was observed, with the exception of 65Cu(n,2n) in the pressure vessel capsule that had a 90% relative difference with respect to the reference results. It is thought that the 65Cu(n,2n) reaction rate reported in the reference for the pressure vessel capsule is not correct. In considering the libraries developed after BUGLE-93, a maximum relative difference of 12% was observed in reaction rates, with respect to the reference results, for 237Np(n,f) in the cavity capsule using BUGLE-B7.

  5. The ANAMMOX reactor under transient-state conditions: process stability with fluctuations of the nitrogen concentration, inflow rate, pH and sodium chloride addition.

    PubMed

    Yu, Jin-Jin; Jin, Ren-Cun

    2012-09-01

    The process stability of an anaerobic ammonium oxidation (ANAMMOX) was investigated in an upflow anaerobic sludge blanket reactor subjected to overloads of 2.0- to 3.0-fold increases in substrate concentrations, inflow rates lasting 12 or 24h, extreme pH levels of 4 and 10 for 12h and a 12-h 30 g l(-1) NaCl addition. During the overloads, the nitrogen removal rate improved, and the shock period was an important factor affecting the reactor performance. In the high pH condition, the reactor performance significantly degenerated; while in the low pH condition, it did not happen. The NaCl addition caused the most serious deterioration in the reactor, which took 108 h to recover and was accompanied by a stoichiometric ratio divergence. There are well correlations between the total nitrogen and the electrical conductivity which is considered to be a convenient signal for controlling and monitoring the ANAMMOX process under transient-state conditions. PMID:22728197

  6. Looking Southwest at Reactor Box Furnaces With Reactor Boxes and ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Looking Southwest at Reactor Box Furnaces With Reactor Boxes and Repossessed Uranium in Recycle Recovery Building - Hematite Fuel Fabrication Facility, Recycle Recovery Building, 3300 State Road P, Festus, Jefferson County, MO

  7. Independent Confirmatory Survey Report for the University of Arizona Nuclear Reactor Laboratory, Tucson, Arizona

    SciTech Connect

    Nick A. Altic

    2011-11-11

    The University of Arizona (University) research reactor is a TRIGA swimming pool type reactor designed by General Atomics and constructed at the University in 1958. The reactor first went into operation in December of 1958 under U.S. Nuclear Regulatory Commission (NRC) license R-52 until final shut down on May 18, 2010. Initial site characterization activities were conducted in February 2009 during ongoing reactor operations to assess the radiological status of the Nuclear Reactor Laboratory (NRL) excluding the reactor tank, associated components, and operating systems. Additional post-shutdown characterization activities were performed to complete characterization activities as well as verify assumptions made in the Decommissioning Plan (DP) that were based on a separate activation analysis (ESI 2009 and WMG 2009). Final status survey (FSS) activities began shortly after the issuance of the FSS plan in May 2011. The contractor completed measurement and sampling activities during the week of August 29, 2011.

  8. 76 FR 62868 - Washington State University; Notice of Issuance of Renewed Facility Operating License No. R-76

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-10-11

    ... published in the Federal Register on June 28, 2010 (75 FR 36717-36721). The NRC received no request for a..., 2011 (76 FR 2192821931), and concluded that renewal of the facility operating license will not have a... Modified TRIGA Nuclear Reactor is a pool-type, natural convection, light-water cooled, and shielded...

  9. Anaerobic digestion of horse dung mixed with different bedding materials in an upflow solid-state (UASS) reactor at mesophilic conditions.

    PubMed

    Böske, Janina; Wirth, Benjamin; Garlipp, Felix; Mumme, Jan; Van den Weghe, Herman

    2014-04-01

    Aim of this study was to investigate the use of upflow anaerobic solid-state (UASS) digestion for treating horse manure. Biochemical methane potential (BMP) tests conducted for varying mixtures of dung (hay and silage feed) and bedding material (wheat straw, flax, hemp, wood chips) showed that straw mixed with hay horse dung has the highest potential of [Formula: see text] . Continuous mesophilic digestion was conducted for 238 days using a single-stage UASS reactor (27 L) and a two-stage UASS system with an anaerobic filter (AF, 21 L). Increasing the organic loading rate (OLR) from 2.5 to 4.5 g vs L(-1)d(-1) enhanced the methane rate of the single-stage reactor from 0.262 to 0.391 LL(-1)d(-1) while the methane yield declined from 104.8 to 86.9 L kg vs(-1). The two-stage system showed similar yields. Thus, for solid-state digestion of horse manure a single-stage UASS reactor appears sufficient. PMID:24583222

  10. McClellan Nuclear Radiation Center (MNRC) TRIGA reactor: The national organization of test research and training reactors

    SciTech Connect

    Kiger, Kevin M.

    1994-07-01

    This year's TRTR conference is being hosted by the McClellan Nuclear Radiation Center. The conference will be held at the Red Lion Hotel in Sacramento, CA. The conference dates are scheduled for October 11-14, 1994. Deadlines for sponsorship commitment and papers have not been set, but are forthcoming. The newly remodeled Red Lion Hotel provides up-to-date conference facilities and one of the most desirable locations for dining, shopping and entertainment in the Sacramento area. While attendees are busy with the conference activities, a spouses program will be available. Although the agenda has not been set, the Sacramento area offers outings to San Francisco, Pier 39, Ghirardelli Square (famous for their chocolate), and a chance to discover 'El Dorado' in the gold country. Not to forget our own bit of history with visits to 'Old Sacramento and Old Folsom', where antiquities abound, to the world renown train museum and incredible eating establishments. (author)

  11. Hybrid systems for transuranic waste transmutation in nuclear power reactors: state of the art and future prospects

    NASA Astrophysics Data System (ADS)

    Yurov, D. V.; Prikhod'ko, V. V.

    2014-11-01

    The features of subcritical hybrid systems (HSs) are discussed in the context of burning up transuranic wastes from the U-Pu nuclear fuel cycle. The advantages of HSs over conventional atomic reactors are considered, and fuel cycle closure alternatives using HSs and fast neutron reactors are comparatively evaluated. The advantages and disadvantages of two HS types with neutron sources (NSs) of widely different natures -- nuclear spallation in a heavy target by protons and nuclear fusion in magnetically confined plasma -- are discussed in detail. The strengths and weaknesses of HSs are examined, and demand for them for closing the U-Pu nuclear fuel cycle is assessed.

  12. Neutronic, steady-state, and transient analyses for the Kazakhstan VVR-K reactor with LEU fuel: ANL independent verification results

    SciTech Connect

    Hanan, Nelson A.; Garner, Patrick L.

    2015-08-01

    Calculations have been performed for steady state and postulated transients in the VVR-K reactor at the Institute of Nuclear Physics (INP), Kazakhstan. (The reactor designation in Cyrillic is BBP-K; transliterating characters to English gives VVR-K but translating words gives WWR-K.) These calculations have been performed at the request of staff of the INP who are performing similar calculations. The selection of the transients considered started during working meetings and email correspondence between Argonne National Laboratory (ANL) and INP staff. In the end the transient were defined by the INP staff. Calculations were performed for the fresh low-enriched uranium (LEU) core and for four subsequent cores as beryllium is added to maintain critically during the first 15 cycles. These calculations have been performed independently from those being performed by INP and serve as one step in the verification process.

  13. Control Rod Malfunction at the NRAD Reactor

    SciTech Connect

    Thomas L. Maddock

    2010-05-01

    The neutron Radiography Reactor (NRAD) is a training, research, and isotope (TRIGA) reactor located at the INL. The reactor is normally shut down by the insertion of three control rods that drop into the core when power is removed from electromagnets. During a routine shutdown, indicator lights on the console showed that one of the control rods was not inserted. It was initially thought that the indicator lights were in error because of a limit switch that was out of adjustment. Through further testing, it was determined that the control rod did not drop when the scram switch was initially pressed. The control rod anomaly led to a six month shutdown of the reactor and an in depth investigation of the reactor protective system. The investigation looked into: scram switch operation, console modifications, and control rod drive mechanisms. A number of latent issues were discovered and corrected during the investigation. The cause of the control rod malfunction was found to be a buildup of corrosion in the control rod drive mechanism. The investigation resulted in modifications to equipment, changes to both operation and maintenance procedures, and additional training. No reoccurrences of the problem have been observed since corrective actions were implemented.

  14. Present status of the medical irradiation facility at the Musashi reactor

    SciTech Connect

    Matsumoto, T.; Aizawa, O.; Nozaki, T.; Sato, T. )

    1989-07-01

    Boron neutron capture therapy (BNCT) of malignant brain tumors has been efficiently performed since March 1977, and the first human case of malignant melanoma was also successfully treated on July 1987 in our reactor (Musashi reactor, TRIGA-II, 100 kW). To obtain both good irradiation field characteristics and a better irradiation facility, some tests and developments have been continued in accordance with the study of medical and biological irradiations. The results of these evaluations and a new approach are presented.6 references.

  15. Digital computer operation of a nuclear reactor

    DOEpatents

    Colley, Robert W.

    1984-01-01

    A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

  16. Digital computer operation of a nuclear reactor

    DOEpatents

    Colley, R.W.

    1982-06-29

    A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

  17. NASA Marshall Space Flight Center Tri-gas Thruster Performance Characterization

    NASA Technical Reports Server (NTRS)

    Dorado, Vanessa; Grunder, Zachary; Schaefer, Bryce; Sung, Meagan; Pedersen, Kevin

    2013-01-01

    Historically, spacecraft reaction control systems have primarily utilized cold gas thrusters because of their inherent simplicity and reliability. However, cold gas thrusters typically have a low specific impulse. It has been determined that a higher specific impulse can be achieved by passing a monopropellant fluid mixture through a catalyst bed prior to expulsion through the thruster nozzle. This research analyzes the potential efficiency improvements from using tri-gas, a mixture of hydrogen, oxygen, and an inert gas, which in this case is helium. Passing tri-gas through a catalyst causes the hydrogen and oxygen to react and form water vapor, ultimately heating the exiting fluid and generating a higher specific impulse. The goal of this project was to optimize the thruster performance by characterizing the effects of varying several system components including catalyst types, catalyst lengths, and initial catalyst temperatures.

  18. Transport model based on three-dimensional cross-section generation for TRIGA core analysis

    SciTech Connect

    Kriangchaiporn, N.; Ivanov, K.; Haghighat, A.; Sears, C. F.

    2006-07-01

    The development of a three-dimensional (3-D) transport model for TRIGA core analysis based on the discrete ordinates (S{sub n}) method has been conducted. The effective fine- and broad- group structures for the TRIGA cross-section libraries were selected based on CPXSD (Contribution and Point-wise Cross-Section Driven) methodology. Different 3-D pin/core configurations are used to verify and validate the selected effective group structures. Thirteen-group structure was finally selected to be used for core analysis. The results agree with continuous energy cross-section Monte Carlo calculations for eigenvalues and normalized pin power distributions, which are used as a reference in this research. (authors)

  19. Prokaryotic diversity and dynamics in a full-scale municipal solid waste anaerobic reactor from start-up to steady-state conditions.

    PubMed

    Cardinali-Rezende, Juliana; Colturato, Luís F D B; Colturato, Thiago D B; Chartone-Souza, Edmar; Nascimento, Andréa M A; Sanz, José L

    2012-09-01

    The prokaryotic diversity of an anaerobic reactor for the treatment of municipal solid waste was investigated over the course of 2 years with the use of 16S rDNA-targeted molecular approaches. The fermentative Bacteroidetes and Firmicutes predominated, and Proteobacteria, Actinobacteria, Tenericutes and the candidate division WWE1 were also identified. Methane production was dominated by the hydrogenotrophic Methanomicrobiales (Methanoculleus sp.) and their syntrophic association with acetate-utilizing and propionate-oxidizing bacteria. qPCR demonstrated the predominance of the hydrogenotrophic over aceticlastic Methanosarcinaceae (Methanosarcina sp. and Methanimicrococcus sp.), and Methanosaetaceae (Methanosaeta sp.) were measured in low numbers in the reactor. According to the FISH and CARD-FISH analyses, Bacteria and Archaea accounted for 85% and 15% of the cells, respectively. Different cell counts for these domains were obtained by qPCR versus FISH analyses. The use of several molecular tools increases our knowledge of the prokaryotic community dynamics from start-up to steady-state conditions in a full-scale MSW reactor. PMID:22750748

  20. STRUCTURAL CALCULATIONS FOR THE CODISPOSAL OF TRIGA SPENT NUCLEAR FUEL IN A WASTE PACKAGE

    SciTech Connect

    S. Mastilovic

    1999-07-28

    The purpose of this analysis is to determine the structural response of a TRIGA Department of Energy (DOE) spent nuclear fuel (SNF) codisposal canister placed in a 5-Defense High Level Waste (DHLW) waste package (WP) and subjected to a tipover design basis event (DBE) dynamic load; the results will be reported in terms of displacements and stress magnitudes. This activity is associated with the WP design.

  1. Decommissioning Plan of the Musashi Reactor and Its Progress

    SciTech Connect

    Tanzawa, Tomio

    2008-01-15

    The Musashi Reactor is a TRIGA-II, tank-type research reactor, as shown in Table 1. The reactor had been operated at maximum thermal power level of 100 kW since first critical, January 30, 1963. Reactor operation was shut down due to small leakage of water from the reactor tank on December 21,1989. After shutdown, investigation of the causes, making plan of repair and discussions on restart or decommissioning had been done. Finally, decision of decommissioning was made in May, 2003. The initial plan of the decommissioning was submitted to the competent authority in January, 2004. Now, the reactor is under decommissioning. The plan of decommissioning and its progress are described. In conclusion: considering the status of undertaking plan of the waste disposal facility for the low level radioactive waste from research reactors, the phased decommissioning was selected for the Musashi Reactor. First phase of the decommissioning activities including the actions of permanent shutdown and delivering the spent nuclear fuels to US DOE was completed.

  2. Participation in the United States Department of Energy Reactor Sharing Program. Annual report, August 31, 1991--August 29, 1992

    SciTech Connect

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1992-05-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics (to become the Department of Mechanical, Aerospace and Nuclear Engineering on July 1, 1992). As such, it is effectively used to support educational programs in engineering and science at the University of Virginia as well as those at other area colleges and universities. The expansion of support to educational programs in the mid-east region is a major objective. To assist in meeting this objective, the University of Virginia has been supported under the US Department of Energy (DOE) Reactor Sharing Program since 1978. Due to the success of the program, this proposal requests continued DOE support through August 1993.

  3. Electrical conductivity as a state indicator for the start-up period of anaerobic fixed-bed reactors.

    PubMed

    Robles, A; Latrille, E; Ribes, J; Bernet, N; Steyer, J P

    2016-01-01

    The aim of this work was to analyse the applicability of electrical conductivity sensors for on-line monitoring the start-up period of an anaerobic fixed-bed reactor. The evolution of bicarbonate concentration and methane production rate was analysed. Strong linear relationships between electrical conductivity and both bicarbonate concentration and methane production rate were observed. On-line estimations of the studied parameters were carried out in a new start-up period by applying simple linear regression models, which resulted in a good concordance between both observed and predicted values. Electrical conductivity sensors were therefore identified as an interesting method for monitoring the start-up period of anaerobic fixed-bed reactors due to its reliability, robustness, easy operation, low cost, and minimum maintenance compared with the currently used sensors. PMID:27148733

  4. Gross Gamma Dose Rate Measurements for TRIGA Spent Nuclear Fuel Burnup Validation

    SciTech Connect

    Winston, Philip Lon; Sterbentz, James William

    2001-04-01

    Gross gamma-ray dose rates from six spent TRIGA fuel elements were measured and compared to calculated values as a means to validate the reported element burnups. A newly installed and functional gamma-ray detection subsystem of the In-Cell Examination System was used to perform the measurements and is described in some detail. The analytical methodology used to calculate the corresponding dose rates is presented along with the calculated values. Comparison of the measured and calculated dose rates for the TRIGA fuel elements indicates good agreement (less than a factor of 2 difference). The intent of the subsystem is to measure the gross gamma dose rate and correlate the measurement to a calculated dose rate based on the element s known burnup and other pertinent spent fuel information. Although validation of the TRIGA elements’ burnup is of primary concern in this paper, the measurement and calculational techniques can be used to either validate an element’s reported burnup or provide a burnup estimate for an element with an unknown burnup.

  5. Gross Gamma Dose Rate Measurements for TRIGA Spent Nuclear Fuel Burnup Validation

    SciTech Connect

    Winston, P.L.; Sterbentz, J.W.

    2002-07-01

    Gross gamma-ray dose rates from six spent TRIGA fuel elements were measured and compared to calculated values as a means to validate the reported element burnups. A newly installed and functional gamma-ray detection subsystem of the In-Cell Examination System was used to perform the measurements and is described in some detail. The analytical methodology used to calculate the corresponding dose rates is presented along with the calculated values. Comparison of the measured and calculated dose rates for the TRIGA fuel elements indicates good agreement (less than a factor of 2 difference). The intent of the subsystem is to measure the gross gamma dose rate and correlate the measurement to a calculated dose rate based on the element s known burnup and other pertinent spent fuel information. Although validation of the TRIGA elements' burnup is of primary concern in this paper, the measurement and calculational techniques can be used to either validate an element's reported burnup or provide a burnup estimate for an element with an unknown burnup. (authors)

  6. Steady-state thermal-hydraulic analysis of the pellet-bed reactor for nuclear thermal propulsion

    SciTech Connect

    El-Genk, M.S.; Morley, N.J.; Yang, J.Y. )

    1992-01-01

    The pellet-bed reactor (PBR) for nuclear thermal propulsion is a hydrogen-cooled, BeO-reflected, fast reactor, consisting of an annular core region filled with randomly packed, spherical fuel pellets. The fuel pellets in the PBR are self-supported, eliminating the need for internal core structure, which simplifies the core design and reduces the size and mass of the reactor. Each spherical fuel pellet is composed of hundreds of fuel microspheres embedded in a zirconium carbide (ZrC) matrix. Each fuel microsphere is composed of a UC-NbC fuel kernel surrounded by two consecutive layers of the NbC and ZrC. Gaseous hydrogen serves both as core coolant and as the propellant for the PBR rocket engine. The cold hydrogen flows axially down the inlet channel situated between the core and the external BeO reflector and radially through the orifices in the cold frit, the core, and the orifices in the hot frit. Finally, the hot hydrogen flows axially out the central channel and exits through converging-diverging nozzle. A thermal-hydraulic analysis of the PBR core was performed with an emphasis on optimizing the size and axial distribution of the orifices in the hot and cold frits to ensure that hot spots would not develop in the core during full-power operation. Also investigated was the validity of the assumptions of neglecting the axial conduction and axial cross flow in the core.

  7. Reactor safety assessment system

    SciTech Connect

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category.

  8. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  9. Pressurized heavy water reactor fuel behaviour in power ramp conditions

    NASA Astrophysics Data System (ADS)

    Ionescu, S.; Uţă, O.; Pârvan, M.; Ohâi, D.

    2009-03-01

    In order to check and improve the quality of the Romanian CANDU fuel, an assembly of six CANDU fuel rods has been subjected to a power ramping test in the 14 MW TRIGA reactor at INR. After testing, the fuel rods have been examined in the hot cells using post-irradiation examination (PIE) techniques such as: visual inspection and photography, eddy current testing, profilometry, gamma scanning, fission gas release and analysis, metallography, ceramography, burn-up determination by mass spectrometry, mechanical testing. This paper describes the PIE results from one out of the six fuel rods. The PIE results concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the sheath, the fission-products activity distribution in the fuel column, the pressure, volume and composition of the fission gas, the burn-up, the isotopic composition and structural changes of the fuel enabled the characterization of the behaviour of the Romanian CANDU fuel in power ramping conditions performed in the TRIGA materials testing reactor.

  10. Studies Related to the Oregon State University High Temperature Test Facility: Scaling, the Validation Matrix, and Similarities to the Modular High Temperature Gas-Cooled Reactor

    SciTech Connect

    Richard R. Schultz; Paul D. Bayless; Richard W. Johnson; William T. Taitano; James R. Wolf; Glenn E. McCreery

    2010-09-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5 year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant project. Because the NRC interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC). Since DOE has incorporated the HTTF as an ingredient in the NGNP thermal-fluids validation program, several important outcomes should be noted: 1. The reference prismatic reactor design, that serves as the basis for scaling the HTTF, became the modular high temperature gas-cooled reactor (MHTGR). The MHTGR has also been chosen as the reference design for all of the other NGNP thermal-fluid experiments. 2. The NGNP validation matrix is being planned using the same scaling strategy that has been implemented to design the HTTF, i.e., the hierarchical two-tiered scaling methodology developed by Zuber in 1991. Using this approach a preliminary validation matrix has been designed that integrates the HTTF experiments with the other experiments planned for the NGNP thermal-fluids verification and validation project. 3. Initial analyses showed that the inherent power capability of the OSU infrastructure, which only allowed a total operational facility power capability of 0.6 MW, is

  11. Test and evaluate the tri-gas low-Btu coal-gasification process. Final report, October 21, 1977-October 31, 1980

    SciTech Connect

    Zabetakis, M.G.

    1980-12-01

    This report describes the continuation of work done to develop the BCR TRI-GAS multiple fluidized-bed gasification process. The objective is the gasification of all ranks of coals with the only product being a clean, low-Btu fuel gas. Design and construction of a 100 lb/h process and equipment development unit (PEDU) was completed on the previous contract. The process consists of three fluid-bed reactors in series, each having a specific function: Stage 1 - pretreatment; Stage 2- - gasification; Stage 3 - maximization of carbon utilization. Under the present contract, 59 PEDU tests have been conducted. A number of these were single-stage tests, mostly in Stage 1; however, integrated PEDU tests were conducted with a western coal (Rosebud) and two eastern coals (Illinois No. 6 and Pittsburgh seam). Both Rosebud and Pittsburgh seam coals were gasified with the PEDU operating in the design mode. Operation with Illinois No. 6 seam coal was also very promising; however, time limitations precluded further testing with this coal. One of the crucial tasks was to operate the Stage 1 reactor to pretreat and devolatilize caking coals. By adding a small amount of air to the fluidizing gas, the caking properties of the coal can be eliminated. However, it was also desirable to release a high percentage of the volatile matter from the coal in this vessel. To accomplish this, the reactor had to be operated above the agglomerating temperature of caking coals. By maintaining a low ratio of fresh to treated coal, this objective was achieved. Both Illinois No. 6 and Pittsburgh seam coals were treated at temperatures of 800 to 900 F without agglomerating in the vessel.

  12. Fuel and core testing plan for a target fueled isotope production reactor.

    SciTech Connect

    Coats, Richard Lee; Dahl, James J.; Parma, Edward J., Jr.

    2010-12-01

    In recent years there has been an unstable supply of the critical diagnostic medical isotope 99Tc. Several concepts and designs have been proposed to produce 99Mo the parent nuclide of 99Tc, at a commercial scale sufficient to stabilize the world supply. This work lays out a testing and experiment plan for a proposed 2 MW open pool reactor fueled by Low Enriched Uranium (LEU) 99Mo targets. The experiments and tests necessary to support licensing of the reactor design are described and how these experiments and tests will help establish the safe operating envelop for a medical isotope production reactor is discussed. The experiments and tests will facilitate a focused and efficient licensing process in order to bring on line a needed production reactor dedicated to supplying medical isotopes. The Target Fuel Isotope Reactor (TFIR) design calls for an active core region that is approximately 40 cm in diameter and 40 cm in fuel height. It contains up to 150 cylindrical, 1-cm diameter, LEU oxide fuel pins clad with Zircaloy (zirconium alloy), in an annular hexagonal array on a {approx}2.0 cm pitch surrounded, radially, by a graphite or a Be reflector. The reactor is similar to U.S. university reactors in power, hardware, and safety/control systems. Fuel/target pin fabrication is based on existing light water reactor fuel fabrication processes. However, as part of licensing process, experiments must be conducted to confirm analytical predictions of steady-state power and accident conditions. The experiment and test plan will be conducted in phases and will utilize existing facilities at the U.S. Department of Energy's Sandia National Laboratories. The first phase is to validate the predicted reactor core neutronics at delayed critical, zero power and very low power. This will be accomplished by using the Sandia Critical Experiment (CX) platform. A full scale TFIR core will be built in the CX and delayed critical measurements will be taken. For low power experiments

  13. Effects of post-reactor functionalization on the phase behaviour of an ethylene-1-octene copolymer studied using solid-state high resolution 13C NMR spectroscopy.

    PubMed

    Calucci, Lucia; Cicogna, Francesca; Forte, Claudia

    2013-10-01

    The effects of post-reactor functionalization with naphthoate-TEMPO on the structure and morphology of an ethylene-1-octene copolymer were investigated by means of solid-state NMR techniques and DSC measurements. Selective (13)C MAS experiments allowed the orthorhombic and the monoclinic crystalline phases and two amorphous phases with different degree of mobility to be detected and quantified. (13)C and (1)H relaxation time measurements and spin diffusion experiments gave insight into the polymer dynamics within the different phases, the crystalline domain dimensions, and the rate of chain diffusion between amorphous and crystalline phases. Comparison of the results obtained for the pristine copolymer and the functionalized samples clearly indicated that the functionalization procedure causes redistribution within the crystalline and the amorphous phases with no relevant change in the degree of crystallinity or in the crystalline domain average size, and slows down chain diffusion. PMID:23942957

  14. BOILING REACTORS

    DOEpatents

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  15. NUCLEAR REACTOR FUEL SYSTEMS

    DOEpatents

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  16. United States Department of Energy`s reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    SciTech Connect

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage.

  17. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

  18. Research reactors

    SciTech Connect

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

    1996-04-01

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

  19. CONVECTION REACTOR

    DOEpatents

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  20. Determination of Unknown Neutron Cross Sections for the Production of Medical Isotopes

    SciTech Connect

    Stephen E. Binney

    2004-04-09

    Calculational assessment and experimental verification of certain neutron cross sections that are related to widely needed new medical isotopes. Experiments were performed at the Oregon State University TRIGA Reactor and the High Flux Irradiation Reactor at Oak Ridge National Laboratory.

  1. Looking Northeast in Oxide Building at Reactors on Second Floor ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Looking Northeast in Oxide Building at Reactors on Second Floor Including Reactor One (Left) and Reactor Two (Right) - Hematite Fuel Fabrication Facility, Oxide Building & Oxide Loading Dock, 3300 State Road P, Festus, Jefferson County, MO

  2. Neutronics, steady-state, and transient analyses for the Poland MARIA reactor for irradiation testing of LEU lead test fuel assemblies from CERCA : ANL independent verification results.

    SciTech Connect

    Garner, P. L.; Hanan, N. A.

    2011-06-07

    The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decide to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process.

  3. Thermal state of the safety system, reactor, side reflector and shielding of the {open_quote}{open_quote}TOPAZ-2{close_quote}{close_quote} system under conditions of fire caused by a launcher accident at the launch pad

    SciTech Connect

    Grinberg, E.I.; Doschatov, V.V.; Nikolaev, V.S.; Sokolov, N.S.; Usov, V.A.

    1996-03-01

    The paper presents some results of calculational analyses performed to determine thermal state of the TOPAZ II safety system structure, radiation shielding, reactor without the side reflector, rods and inserts of the side reflector under conditions of fire at the launch pad when an accident occurs to a launch vehicle. {copyright} {ital 1996 American Institute of Physics.}

  4. Radiological consequences of ship collisions that might occur in U.S. Ports during the shipment of foreign research reactor spent nuclear fuel to the United States in break-bulk freighters

    SciTech Connect

    Sprung, J.L.; Bespalko, S.J.; Massey, C.D.; Yoshimura, R.; Johnson, J.D.; Reardon, P.C.; Ebert, M.W.; Gallagher D.W.

    1996-08-01

    Accident source terms, source term probabilities, consequences, and risks are developed for ship collisions that might occur in U.S. ports during the shipment of spent fuel from foreign research reactors to the United States in break-bulk freighters.

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

    NASA Astrophysics Data System (ADS)

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

    2008-03-01

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

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

    SciTech Connect

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

    2008-03-17

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

  7. A pyrosequencing-based metagenomic study of methane-producing microbial community in solid-state biogas reactor

    PubMed Central

    2013-01-01

    Background A solid-state anaerobic digestion method is used to produce biogas from various solid wastes in China but the efficiency of methane production requires constant improvement. The diversity and abundance of relevant microorganisms play important roles in methanogenesis of biomass. The next-generation high-throughput pyrosequencing platform (Roche/454 GS FLX Titanium) provides a powerful tool for the discovery of novel microbes within the biogas-generating microbial communities. Results To improve the power of our metagenomic analysis, we first evaluated five different protocols for extracting total DNA from biogas-producing mesophilic solid-state fermentation materials and then chose two high-quality protocols for a full-scale analysis. The characterization of both sequencing reads and assembled contigs revealed that the most prevalent microbes of the fermentation materials are derived from Clostridiales (Firmicutes), which contribute to degrading both protein and cellulose. Other important bacterial species for decomposing fat and carbohydrate are Bacilli, Gammaproteobacteria, and Bacteroidetes (belonging to Firmicutes, Proteobacteria, and Bacteroidetes, respectively). The dominant bacterial species are from six genera: Clostridium, Aminobacterium, Psychrobacter, Anaerococcus, Syntrophomonas, and Bacteroides. Among them, abundant Psychrobacter species, which produce low temperature-adaptive lipases, and Anaerococcus species, which have weak fermentation capabilities, were identified for the first time in biogas fermentation. Archaea, represented by genera Methanosarcina, Methanosaeta and Methanoculleus of Euryarchaeota, constitute only a small fraction of the entire microbial community. The most abundant archaeal species include Methanosarcina barkeri fusaro, Methanoculleus marisnigri JR1, and Methanosaeta theromphila, and all are involved in both acetotrophic and hydrogenotrophic methanogenesis. Conclusions The identification of new bacterial genera and

  8. Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements

    SciTech Connect

    J. D. Bess; T. L. Maddock; M. A. Marshall

    2011-09-01

    The Neutron Radiography (NRAD) reactor is a 250-kW TRIGA-(Training, Research, Isotope Production, General Atomics)-conversion-type reactor at the Idaho National Laboratory; it is primarily used for neutron radiography analysis of irradiated and unirradiated fuels and materials. The NRAD reactor was converted from HEU to LEU fuel with 60 fuel elements and brought critical on March 31, 2010. This configuration of the NRAD reactor has been evaluated as an acceptable benchmark experiment and is available in the 2011 editions of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Significant effort went into precisely characterizing all aspects of the reactor core dimensions and material properties; detailed analyses of reactor parameters minimized experimental uncertainties. The largest contributors to the total benchmark uncertainty were the 234U, 236U, Er, and Hf content in the fuel; the manganese content in the stainless steel cladding; and the unknown level of water saturation in the graphite reflector blocks. A simplified benchmark model of the NRAD reactor was prepared with a keff of 1.0012 {+-} 0.0029 (1s). Monte Carlo calculations with MCNP5 and KENO-VI and various neutron cross section libraries were performed and compared with the benchmark eigenvalue for the 60-fuel-element core configuration; all calculated eigenvalues are between 0.3 and 0.8% greater than the benchmark value. Benchmark evaluations of the NRAD reactor are beneficial in understanding biases and uncertainties affecting criticality safety analyses of storage, handling, or transportation applications with LEU-Er-Zr-H fuel.

  9. Reactor Opportunities for the Spheromak

    SciTech Connect

    Hooper, E B; Bulmer, R H; Fowler, T K; Hill, D N; McLean, H S; Romero-Talamas, C A; Moir, R W; Stallard, B W; Wood, R D; Woodruff, S

    2003-04-18

    Experimental results from the Sustained Spheromak Physics Experiment, SSPX, are reviewed and applied to published reactor configurations. The results include several important features, including low fluctuation levels, (apparent) good magnetic flux surfaces, and moderate beta. Additional features needed for an attractive reactor but not yet demonstrated experimentally are identified by comparison with the reactor designs, and possible alternatives to a fully steady-state device are discussed.

  10. NEUTRONIC REACTOR

    DOEpatents

    Fraas, A.P.; Mills, C.B.

    1961-11-21

    A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

  11. REACTOR COOLING

    DOEpatents

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  12. A CAMAC based real-time noise analysis system for nuclear reactors

    NASA Astrophysics Data System (ADS)

    Ciftcioglu, Özer

    1987-05-01

    A CAMAC based real-time noise analysis system was designed for the TRIGA MARK II nuclear reactor at the Institute for Nuclear Energy, Istanbul. The input analog signals obtained from the radiation detectors are introduced to the system through CAMAC interface. The signals converted into digital form are processed by a PDP-11 computer. The fast data processing based on auto/cross power spectral density computations is carried out by means of assembly written FFT algorithms in real-time and the spectra obtained are displayed on a CAMAC driven display system as an additional monitoring device. The system has the advantage of being software programmable and controlled by a CAMAC system so that it is operated under program control for reactor surveillance, anomaly detection and diagnosis. The system can also be used for the identification of nonstationary operational characteristics of the reactor in long term by comparing the noise power spectra with the corresponding reference noise patterns prepared in advance.

  13. Optimally moderated nuclear fission reactor and fuel source therefor

    DOEpatents

    Ougouag, Abderrafi M.; Terry, William K.; Gougar, Hans D.

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  14. 75 FR 36717 - Washington State University; Notice of Acceptance for Docketing and Opportunity for Hearing on...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-28

    ... University (the licensee, WSU) to operate the Washington State University Modified TRIGA Nuclear Radiation... NRC E Filing rule (72 FR 49139, August 28, 2007). The E Filing process requires participants to submit... NRC's E Filing Rule (72 FR 49139; August 28, 2007) apply to appeals of NRC staff...

  15. Breazeale Reactor Modernization Program

    SciTech Connect

    Davison, C. C.

    2003-04-16

    The Penn State Breazeale Nuclear Reactor is the longest operating licensed research reactor in the nation. The facility has played a key role in educating scientists, engineers and in providing facilities and services to researchers in many different disciplines. In order to remain a viable and effective research and educational institution, a multi-phase modernization project was proposed. Phase I was the replacement of the 25-year old reactor control and safety system along with associated wiring and hardware. This phase was fully funded by non-federal funds. Tasks identified in Phases II-V expand upon and complement the work done in Phase I to strategically implement state-of-the-art technologies focusing on identified national needs and priorities of the future.

  16. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1958-04-22

    A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

  17. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.; Johnson, H.W.

    1961-04-01

    BS>A nuclear reactor incorporating fuel rods passing through a moderator and including tubes of a material of higher Thermal conductivity than the fuel in contact with the fuel is described. The tubes extend beyond the active portion of the reactor into contant with a fiuld coolant.

  18. F Reactor Inspection

    SciTech Connect

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-10-29

    Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

  19. F Reactor Inspection

    ScienceCinema

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-11-24

    Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

  20. Accuracy studies with carbon clusters at the Penning trap mass spectrometer TRIGA-TRAP

    NASA Astrophysics Data System (ADS)

    Ketelaer, J.; Beyer, T.; Blaum, K.; Block, M.; Eberhardt, K.; Eibach, M.; Herfurth, F.; Smorra, C.; Nagy, Sz.

    2010-05-01

    Extensive cross-reference measurements of well-known frequency ratios using various sizes of carbon cluster ions 12Cn + (10≤n≤23) were performed to determine the effects limiting the accuracy of mass measurements at the Penning-trap facility TRIGA-TRAP. Two major contributions to the uncertainty of a mass measurement have been identified. Fluctuations of the magnetic field cause an uncertainty in the frequency ratio due to the required calibration by a reference ion of uf(νref)/νref = 6(2) × 10-11/min × Δt. A mass-dependent systematic shift of the frequency ratio of epsilonm(r)/r = -2.2(2) × 10-9 × (m-mref)/u has been found as well. Finally, the nuclide 197Au was used as a cross-check since its mass is already known with an uncertainty of 0.6 keV.

  1. Final qualification testing results for TRIGA low-enriched uranium fuel

    SciTech Connect

    West, G.B.; Simnad, M.T.; Copeland, G.L.

    1987-01-01

    Following the adoption of policies by the US government and other fuel supplier nations to limit, with few exceptions, export of fuels to those enriched to < 20% in /sup 235/U, GA Technologies undertook, starting in 1976, a rigorous program to develop uranium zirconium-hydride fuels with high uranium concentrations up to 3.7 g U/cm/sup 3/, while limiting the enrichment to < 20% in /sup 235/U. By increasing the uranium concentration from 8.5 to 45 wt% and by including erbium as a burnable poison, the long reactivity lifetime of the high-enriched uranium cores has been preserved. The achieved purpose of these low-enriched uranium (LEU) fuels was also to preserve all the unique safety features of the TRIGA fuel system - prompt negative temperature coefficient of reactivity, high fission product retentivity, chemical stability when quenched from high temperatures in water, and dimensional stability over a broad range of operating temperatures.

  2. The Optimum Plutonium Fuel Form in Light Water Reactors

    NASA Astrophysics Data System (ADS)

    Tulenko, James S.; Savela, Michael; Gueorguiev, Gueorgui

    2003-07-01

    The University of Florida has underway a research program to validate the benefits of developing a Pu/ZrH/U matrix fuel for the irradiation of the U.S. weapons plutonium and European reprocessed plutonium from an economic, operational, and performance basis. Thermal reactors using plutonium as a fuel are inherently undermoderated because of the large absorption cross sections of plutonium and the presence of large absorption resonances for plutonium in the thermal and near-thermal energy ranges. The use of the proven TRIGA ZrHx-based fuel with plutonium has shown an extremely large (>20%) increase in reactivity over the conventional UO2/PuO2 fuel form currently being considered, with an additional major increase in the destruction of plutonium, rendering it an extremely attractive fuel form for plutonium disposition.

  3. Drive reinforcement neural networks for reactor control. Final report

    SciTech Connect

    Williams, J.G.; Jouse, W.C.

    1995-02-01

    In view of the loss of the third year funding, the scope of the project goals has been revised. The revision in project scope no longer allows for the detailed modeling of the EBR-11 start-up task that was originally envisaged. The authors are continuing, however, to model the control of the rapid power ascent of the University of Arizona TRIGA reactor using a model-based controller and using a drive reinforcement neural network. These will be combined during the concluding period of the project into a hierarchical control architecture. In addition, the modeling of a PWR feedwater heater has continued, and an autonomous fault-tolerant software architecture for its control has been proposed.

  4. NUCLEAR REACTOR

    DOEpatents

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  5. Plug Flow Reactor Simulator

    SciTech Connect

    Larson, Richard S.

    1996-07-30

    PLUG is a computer program that solves the coupled steady state continuity, momentum, energy, and species balance equations for a plug flow reactor. Both homogeneous (gas-phase) and heterogenous (surface) reactions can be accommodated. The reactor may be either isothermal or adiabatic or may have a specified axial temperature or heat flux profile; alternatively, an ambient temperature and an overall heat-transfer coefficient can be specified. The crosssectional area and surface area may vary with axial position, and viscous drag is included. Ideal gas behavior and surface site conservation are assumed.

  6. Plug Flow Reactor Simulator

    Energy Science and Technology Software Center (ESTSC)

    1996-07-30

    PLUG is a computer program that solves the coupled steady state continuity, momentum, energy, and species balance equations for a plug flow reactor. Both homogeneous (gas-phase) and heterogenous (surface) reactions can be accommodated. The reactor may be either isothermal or adiabatic or may have a specified axial temperature or heat flux profile; alternatively, an ambient temperature and an overall heat-transfer coefficient can be specified. The crosssectional area and surface area may vary with axial position,more » and viscous drag is included. Ideal gas behavior and surface site conservation are assumed.« less

  7. NEUTRONIC REACTOR

    DOEpatents

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  8. NEUTRONIC REACTOR

    DOEpatents

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  9. NUCLEAR REACTOR

    DOEpatents

    Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

    1962-10-23

    A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

  10. REACTOR SHIELD

    DOEpatents

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  11. NUCLEAR REACTOR

    DOEpatents

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  12. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  13. NUCLEAR REACTOR

    DOEpatents

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  14. NUCLEAR REACTOR

    DOEpatents

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

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

    SciTech Connect

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

    2013-07-01

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

  16. 77 FR 7613 - Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-02-13

    ... provides text and image files of the NRC's public documents. If you do not have access to ADAMS or if there... in the hearing. For each contention, the petitioner must provide a specific statement of the issue of... contention. Additionally, the petitioner must demonstrate that the issue raised by each contention is...

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

    SciTech Connect

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

    1996-05-01

    As part of the Department of Energy`s spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, are not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report.

  18. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  19. Thermionic Reactor Design Studies

    SciTech Connect

    Schock, Alfred

    1994-08-01

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  20. Studies of the behavior of a reactor neutron beam at the sample position of a diffractometer using silicon monochromators

    NASA Astrophysics Data System (ADS)

    Ahmed, F. U.; Ahsan, M. H.; Khan, Aysha A.; Kamal, I.; Awal, M. A.; Ahmad, A. A. Z.

    1992-02-01

    A computer program TISTA has been developed for calculation of different aspects of designing a double axis neutron spectrometer at the TRIGA Mark II research reactor of the Atomic Energy Research Establishment, Dhaka, Bangladesh. The mathematical algorithms used in this program are based on the formalisms used by Fischer, Sabine and Bacon. Angle and energy resolutions and flux density as functions of neutron wave length, beam collimation, crystal asymmetry and deviation from zero-Bragg-angle position for different silicon crystal planes (111, 220, 311) have been calculated.

  1. NEUTRONIC REACTOR

    DOEpatents

    Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

    1958-09-01

    This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

  2. Stabilized Spheromak Fusion Reactors

    SciTech Connect

    Fowler, T

    2007-04-03

    The U.S. fusion energy program is focused on research with the potential for studying plasmas at thermonuclear temperatures, currently epitomized by the tokamak-based International Thermonuclear Experimental Reactor (ITER) but also continuing exploratory work on other plasma confinement concepts. Among the latter is the spheromak pursued on the SSPX facility at LLNL. Experiments in SSPX using electrostatic current drive by coaxial guns have now demonstrated stable spheromaks with good heat confinement, if the plasma is maintained near a Taylor state, but the anticipated high current amplification by gun injection has not yet been achieved. In future experiments and reactors, creating and maintaining a stable spheromak configuration at high magnetic field strength may require auxiliary current drive using neutral beams or RF power. Here we show that neutral beam current drive soon to be explored on SSPX could yield a compact spheromak reactor with current drive efficiency comparable to that of steady state tokamaks. Thus, while more will be learned about electrostatic current drive in coming months, results already achieved in SSPX could point to a productive parallel development path pursuing auxiliary current drive, consistent with plans to install neutral beams on SSPX in the near future. Among possible outcomes, spheromak research could also yield pulsed fusion reactors at lower capital cost than any fusion concept yet proposed.

  3. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  4. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2013-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has been evaluated as an acceptable benchmark experiment. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  5. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2011-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  6. The Effect of Pitch, Burnup, and Absorbers on a TRIGA Spent-Fuel Pool Criticality Safety

    SciTech Connect

    Logar, Marjan; Jeraj, Robert; Glumac, Bogdan

    2003-02-15

    It has been shown that supercriticality might occur for some postulated accident conditions at the TRIGA spent-fuel pool. However, the effect of burnup was not accounted for in previous studies. In this work, the combined effect of fuel burnup, pitch among fuel elements, and number of uniformly mixed absorber rods for a square arrangement on the spent-fuel pool k{sub eff} is investigated.The Monte Carlo computer code MCNP4B with the ENDF-B/VI library and detailed three dimensional geometry was used. The WIMS-D code was used to model the isotopic composition of the standard TRIGA and FLIP fuel for 5, 10, 20 and 30% burnup level and 2- and 4-yr cooling time.The results show that out of the three studied effects, pitch from contact (3.75 cm) up to rack design pitch (8 cm), number of absorbers from zero to eight, and burnup up to 30%, the pitch has the greatest influence on the multiplication factor k{sub eff}. In the interval in which the pitch was changed, k{sub eff} decreased for up to {approx}0.4 for standard and {approx}0.3 for FLIP fuel. The number of absorber rods affects the multiplication factor much less. This effect is bigger for more compact arrangements, e.g., for contact of standard fuel elements with eight absorber rods among them, k{sub eff} values are smaller for {approx}0.2 ({approx}0.1 for FLIP) than for arrangements without absorber rods almost regardless of the burnup. The effect of burnup is the smallest. For standard fuel elements, it is {approx}0.1 for almost all pitches and numbers of absorbers. For FLIP fuel, it is smaller for a factor of 3, but increases with the burnup for compact arrangements. Cooling time of fuel has just a minor effect on the k{sub eff} of spent-fuel pool and can be neglected in spent-fuel pool design.

  7. Challenges in the Development of Advanced Reactors

    SciTech Connect

    P. Sabharwall; M.C. Teague; S.M. Bragg-Sitton; M.W. Patterson

    2012-08-01

    Past generations of nuclear reactors have been successively developed and the next generation is currently being developed, demonstrating the constant progress and technical and industrial vitality of nuclear energy. In 2000 US Department of Energy launched Generation IV International Forum (GIF) which is one of the main international frameworks for the development of future nuclear systems. The six systems that were selected were: sodium cooled fast reactor, lead cooled fast reactor, supercritical water cooled reactor, very high temperature gas cooled reactor (VHTR), gas cooled fast reactor and molten salt reactor. This paper discusses some of the proposed advanced reactor concepts that are currently being researched to varying degrees in the United States, and highlights some of the major challenges these concepts must overcome to establish their feasibility and to satisfy licensing requirements.

  8. NUCLEAR REACTOR

    DOEpatents

    Christy, R.F.

    1958-07-15

    A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

  9. POWER REACTOR

    DOEpatents

    Zinn, W.H.

    1958-07-01

    A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

  10. NEUTRONIC REACTORS

    DOEpatents

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  11. Catalytic reactor

    DOEpatents

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  12. Bioconversion reactor

    SciTech Connect

    McCarty, P.L.; Bachmann, A.

    1992-02-25

    A bioconversion reactor is described for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible. 7 figs.

  13. Bioconversion reactor

    DOEpatents

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  14. REACTOR CONTROL

    DOEpatents

    Fortescue, P.; Nicoll, D.

    1962-04-24

    A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

  15. NUCLEAR REACTOR

    DOEpatents

    Young, G.

    1963-01-01

    This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

  16. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  17. Neutronic reactor

    DOEpatents

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  18. Neutronic reactor

    DOEpatents

    Wende, Charles W. J.

    1976-08-17

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

  19. Sonochemical Reactors.

    PubMed

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation. PMID:27573503

  20. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1960-09-27

    A unit assembly is described for a neutronic reactor comprising a tube and plurality of spaced parallel sandwiches in the tube extending lengthwise thereof, each sandwich including a middle plate having a central opening for plutonium and other openings for fertile material at opposite ends of the plate.

  1. NEUTRONIC REACTOR

    DOEpatents

    Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

    1959-10-27

    BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

  2. Neutronic reactor

    DOEpatents

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  3. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect

    R. Wigeland; J. Cahalan

    2009-09-01

    thermophysical properties of the fuel and their compatibility with the reactor coolant, with corresponding differences in the challenges presented to the reactor developers. Accident phenomena are discussed for the sodium-cooled fast reactor based on the mechanistic progression of conditions from accident initiation to accident termination, whether a benign state is achieved or more severe consequences are expected. General principles connecting accident phenomena and fuel properties are developed from the oxide and metal fuel safety analyses, providing guidelines that can be used as part of the evaluation for selection of fuel type for the sodium-cooled fast reactor.

  4. Dose Calculations for the Co-Disposal WP-of HLW-Glass and the Triga SNF

    SciTech Connect

    G. Radulescu

    1999-08-02

    This calculation is prepared by the Monitored Geologic Repository (MGR) Waste Package Operations (WPO). The purpose of this calculation is to determine the surface dose rates of a codisposal waste package (WP) containing a centrally located Department of Energy (DOE) standardized 18-in. spent nuclear fuel (SNF) canister, loaded with the TRIGA (Training, Research, Isotopes, General Atomics) SNF. This canister is surrounded by five 3-m long canisters, loaded with Savannah River Site (SRS) high-level waste (HLW) glass. The results are to support the WP design and radiological analyses.

  5. NEUTRONIC REACTOR

    DOEpatents

    Wade, E.J.

    1958-09-16

    This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

  6. REACTOR MONITORING

    DOEpatents

    Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

    1959-02-01

    A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

  7. REACTOR UNLOADING

    DOEpatents

    Leverett, M.C.

    1958-02-18

    This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

  8. Nuclear reactor

    DOEpatents

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  9. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1958-08-19

    A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

  10. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  11. NUCLEAR REACTORS

    DOEpatents

    Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

    1961-12-01

    An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

  12. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  13. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1961-01-24

    A core structure for neutronic reactors adapted for the propulsion of aircraft and rockets is offered. The core is designed for cooling by gaseous media, and comprises a plurality of hollow tapered tubular segments of a porous moderating material impregniated with fissionable fuel nested about a common axis. Alternate ends of the segments are joined. In operation a coolant gas passes through the porous structure and is heated.

  14. REACTOR CONTROL

    DOEpatents

    Ruano, W.J.

    1957-12-10

    This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

  15. BOILER-SUPERHEATED REACTOR

    DOEpatents

    Heckman, T.P.

    1961-05-01

    A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

  16. Ultrasonic inspection of reactor systems

    SciTech Connect

    Majzlik, E.J. Jr.

    1989-01-01

    The subject of this presentation is ultrasonic inspection of reactor systems. This paper describes two current programs underway at Savannah River Site which provide state-of-the-art ultrasonic inspections of weld heat-affected zones in the primary cooling loop of the Savannah River Site reactors. It also describes the automated remote inspection equipment being developed and employed; briefly describe the procedures being used; and give you a general idea of the future direction of two major programs: Moderator Piping Inspection Program and the Reactor Tank Wall Weld Inspection Program. The objective of these programs is to provide inspection techniques to more fully determine the condition of the reactor primary system and provide data for prediction of maintenance needs and remaining service life. Detection and sizing of intergranular stress corrosion cracking is the focus of these programs.

  17. SUSEE: A Compact, Lightweight Space Nuclear Power System Using Present Water Reactor Technology

    NASA Astrophysics Data System (ADS)

    Maise, George; Powell, James; Paniagua, John

    2006-01-01

    The SUSEE space reactor system uses existing nuclear fuels and the standard steam cycle to generate electrical and thermal power for a wide range of in-space and surface applications, including manned bases, sub-surface mobile probes to explore thick ice deposits on Mars and the Jovian moons, and mobile rovers. SUSEE cycle efficiency, thermal to electric, ranges from ~20 to 24%, depending on operating parameters. Rejection of waste heat is by a lightweight condensing radiator that can be launched as a compact rolled-up package and deployed into flat panels when appropriate. The 50 centimeter diameter SUSEE reactor can provide power over the range of 10 kW(e) to 1 MW(e) for a period of 10 years. Higher power outputs are possible using slightly larger reactors. System specific weight (reactor, turbine, generator, piping, and radiator is ~3 kg/kW(e). Two SUSEE reactor options are described, based on the existing Zr/O2 cermet and the UH3/ZrH2 TRIGA nuclear fuels.

  18. The rate of decay of fresh fission products from a nuclear reactor

    NASA Astrophysics Data System (ADS)

    Dolan, David J.

    Determining the rate of decay of fresh fission products from a nuclear reactor is complex because of the number of isotopes involved, different types of decay, half-lives of the isotopes, and some isotopes decay into other radioactive isotopes. Traditionally, a simplified rule of 7s and 10s is used to determine the dose rate from nuclear weapons and can be to estimate the dose rate from fresh fission products of a nuclear reactor. An experiment was designed to determine the dose rate with respect to time from fresh fission products of a nuclear reactor. The experiment exposed 0.5 grams of unenriched Uranium to a fast and thermal neutron flux from a TRIGA Research Reactor (Lakewood, CO) for ten minutes. The dose rate from the fission products was measured by four Mirion DMC 2000XB electronic personal dosimeters over a period of six days. The resulting dose rate following a rule of 10s: the dose rate of fresh fission products from a nuclear reactor decreases by a factor of 10 for every 10 units of time.

  19. SUSEE: A Compact, Lightweight Space Nuclear Power System Using Present Water Reactor Technology

    SciTech Connect

    Maise, George; Powell, James; Paniagua, John

    2006-01-20

    The SUSEE space reactor system uses existing nuclear fuels and the standard steam cycle to generate electrical and thermal power for a wide range of in-space and surface applications, including manned bases, sub-surface mobile probes to explore thick ice deposits on Mars and the Jovian moons, and mobile rovers. SUSEE cycle efficiency, thermal to electric, ranges from {approx}20 to 24%, depending on operating parameters. Rejection of waste heat is by a lightweight condensing radiator that can be launched as a compact rolled-up package and deployed into flat panels when appropriate. The 50 centimeter diameter SUSEE reactor can provide power over the range of 10 kW(e) to 1 MW(e) for a period of 10 years. Higher power outputs are possible using slightly larger reactors. System specific weight (reactor, turbine, generator, piping, and radiator) is {approx}3 kg/kW(e). Two SUSEE reactor options are described, based on the existing Zr/O2 cermet and the UH3/ZrH2 TRIGA nuclear fuels.

  20. Small Reactor for Deep Space Exploration

    SciTech Connect

    none,

    2012-11-29

    This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

  1. Small Reactor for Deep Space Exploration

    ScienceCinema

    none,

    2014-05-30

    This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

  2. Nuclear reactor

    DOEpatents

    Pennell, William E.; Rowan, William J.

    1977-01-01

    A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

  3. ELECTRONUCLEAR REACTOR

    DOEpatents

    Lawrence, E.O.; McMillan, E.M.; Alvarez, L.W.

    1960-04-19

    An electronuclear reactor is described in which a very high-energy particle accelerator is employed with appropriate target structure to produce an artificially produced material in commercial quantities by nuclear transformations. The principal novelty resides in the combination of an accelerator with a target for converting the accelerator beam to copious quantities of low-energy neutrons for absorption in a lattice of fertile material and moderator. The fertile material of the lattice is converted by neutron absorption reactions to an artificially produced material, e.g., plutonium, where depleted uranium is utilized as the fertile material.

  4. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashley, J.W.

    1958-12-16

    A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

  5. REACTOR COMPONETN

    DOEpatents

    Creutz, E.C.

    1959-10-27

    A reactor fuel element comprised of a slug of fissionable material disposed in a sheath of corrosion resistantmaterial is described. The sheath is in the form of a tubular container closed at one end and is in tight-fitting engagement with the peripheral sunface of the slug. An inner cap is insented into the open end of the sheath against the slug, which end is then bent around the inner cap and welded thereto. An outer cap is then welded around its peripheny to the bent portion of the container.

  6. Cooperative Monitoring of Reactors with Antineutrino Detectors

    NASA Astrophysics Data System (ADS)

    Keefer, Greg

    2010-11-01

    The current state-of-the-art in antineutrino detection is such that it is now possible to monitor the operational status, power levels and fissile content of nuclear reactors in real-time at standoff distances of a few tens of meters, well outside of the reactor containment. This has been demonstrated at civilian power reactors in both Russia and the United States. In the last few years, the International Atomic Energy Agency has begun to consider the potential of this technology for its reactor safeguards regime. In this talk, I describe the state of the art for this application, and emphasize the natural overlap with ongoing efforts in fundamental physics to measure the oscillations of antineutrinos using reactor sources.

  7. Helium-cooled high temperature reactors

    SciTech Connect

    Trauger, D.B.

    1985-01-01

    Experience with several helium cooled reactors has been favorable, and two commercial plants are now operating. Both of these units are of the High Temperature Graphite Gas Cooled concept, one in the United States and the other in the Federal Republic of Germany. The initial helium charge for a reactor of the 1000 MW(e) size is modest, approx.15,000 kg.

  8. Hybrid adsorptive membrane reactor

    DOEpatents

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  9. Hybrid adsorptive membrane reactor

    NASA Technical Reports Server (NTRS)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  10. Control Means for Reactor

    DOEpatents

    Manley, J. H.

    1961-06-27

    An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

  11. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1962-12-18

    A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

  12. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1957-09-17

    A reactor of the type having coolant liquid circulated through clad fuel elements geometrically arranged in a solid moderator, such as graphite, is described. The core is enclosed in a pressure vessel and suitable shielding, wherein means is provided for circulating vapor through the core to superheat the same. This is accomplished by drawing off the liquid which has been heated in the core due to the fission of the fuel, passing it to a nozzle within a chamber where it flashes into a vapor, and then passing the vapor through separate tubes extending through the moderator to pick up more heat developed in the core due to the fission of the fuel, thereby producing superheated vapor.

  13. NEUTRONIC REACTOR

    DOEpatents

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  14. Advanced fuels modeling: Evaluating the steady-state performance of carbide fuel in helium-cooled reactors using FRAPCON 3.4

    NASA Astrophysics Data System (ADS)

    Hallman, Luther, Jr.

    Uranium carbide (UC) has long been considered a potential alternative to uranium dioxide (UO2) fuel, especially in the context of Gen IV gas-cooled reactors. It has shown promise because of its high uranium density, good irradiation stability, and especially high thermal conductivity. Despite its many benefits, UC is known to swell at a rate twice that of UO2. However, the swelling phenomenon is not well understood, and we are limited to a weak empirical understanding of the swelling mechanism. One suggested cladding for UC is silicon carbide (SiC), a ceramic that demonstrates a number of desirable properties. Among them are an increased corrosion resistance, high mechanical strength, and irradiation stability. However, with increased temperatures, SiC exhibits an extremely brittle nature. The brittle behavior of SiC is not fully understood and thus it is unknown how SiC would respond to the added stress of a swelling UC fuel. To better understand the interaction between these advanced materials, each has been implemented into FRAPCON, the preferred fuel performance code of the Nuclear Regulatory Commission (NRC); additionally, the material properties for a helium coolant have been incorporated. The implementation of UC within FRAPCON required the development of material models that described not only the thermophysical properties of UC, such as thermal conductivity and thermal expansion, but also models for the swelling, densification, and fission gas release associated with the fuel's irradiation behavior. This research is intended to supplement ongoing analysis of the performance and behavior of uranium carbide and silicon carbide in a helium-cooled reactor.

  15. Nuclear reactor

    DOEpatents

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  16. Performance of Magnet Insulation Systems at Low Temperature and After Reactor Irradiation

    SciTech Connect

    Bittner-Rohrhofer, K.; Humer, K.; Fillunger, H.; Maix, R.K.; Weber, H.W.

    2004-06-28

    Advanced composite materials reinforced with boron-free glass fibers are candidate insulation materials for fusion magnets, in particular for ITER. Thus, these systems require an excellent performance and mechanical integrity after irradiation. The present innovative organic insulation system consists of R-glass fiber reinforced tapes impregnated with an advanced cyanate-ester/epoxy resin. This composite is suitable for vacuum-pressure impregnation. In order to assess the radiation resistance of the mechanical properties, the laminate was irradiated in the TRIGA reactor (Vienna) to the ITER design fluence level of 1x1022 m-2 (E>0.1 MeV). The blend was screened at 77 K using the static tensile and short-beam-shear test prior to and after irradiation. In addition, tension-tension fatigue measurements were done in order to investigate the material performance under pulsed operating conditions.

  17. Performance of Magnet Insulation Systems at Low Temperature and After Reactor Irradiation

    NASA Astrophysics Data System (ADS)

    Bittner-Rohrhofer, K.; Humer, K.; Fillunger, H.; Maix, R. K.; Weber, H. W.

    2004-06-01

    Advanced composite materials reinforced with boron-free glass fibers are candidate insulation materials for fusion magnets, in particular for ITER. Thus, these systems require an excellent performance and mechanical integrity after irradiation. The present innovative organic insulation system consists of R-glass fiber reinforced tapes impregnated with an advanced cyanate-ester/epoxy resin. This composite is suitable for vacuum-pressure impregnation. In order to assess the radiation resistance of the mechanical properties, the laminate was irradiated in the TRIGA reactor (Vienna) to the ITER design fluence level of 1×1022 m-2 (E>0.1 MeV). The blend was screened at 77 K using the static tensile and short-beam-shear test prior to and after irradiation. In addition, tension-tension fatigue measurements were done in order to investigate the material performance under pulsed operating conditions.

  18. Preparation and quantification of radioactive particles for tracking hydrodynamic behavior in multiphase reactors.

    PubMed

    Yunos, Mohd Amirul Syafiq Mohd; Hussain, Siti Aslina; Yusoff, Hamdan Mohamed; Abdullah, Jaafar

    2014-09-01

    Radioactive particle tracking (RPT) has emerged as a promising and versatile technique that can provide rich information about a variety of multiphase flow systems. However, RPT is not an off-the-shelf technique, and thus, users must customize RPT for their applications. This paper presents a simple procedure for preparing radioactive tracer particles created via irradiation with neutrons from the TRIGA Mark II research reactor. The present study focuses on the performance evaluation of encapsulated gold and scandium particles for applications as individual radioactive tracer particles using qualitative and quantitative neutron activation analysis (NAA) and an X-ray microcomputed tomography (X-ray Micro-CT) scanner installed at the Malaysian Nuclear Agency. PMID:24907683

  19. Reactor and method of operation

    DOEpatents

    Wheeler, John A.

    1976-08-10

    A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

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

    SciTech Connect

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

    2008-07-15

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

  1. Toward reactor monitoring with antineutrinos

    SciTech Connect

    Guillon, Benoit; Cormon, S.; Fallot, M.; Giot, L.; Martino, J.; Cribier, M.; Lasserre, T.

    2007-07-01

    The fundamental knowledge on neutrino properties acquired in recent years as well as the great experimental progress made on neutrino detection open nowadays the possibility of applied neutrino physics. Among it, the International Atomic Energy Agency (IAEA) asked to its member states to study the possibility of nuclear reactor monitoring applications, such as the thermal power measurement or the fuel composition bookkeeping. In this context, we report studies aiming at a better determination of the antineutrino energy spectrum emitted by nuclear power plants, necessary for reactor monitoring applications, but also for experiments studying the ground properties of these particles. (authors)

  2. Reactor safety method

    DOEpatents

    Vachon, Lawrence J.

    1980-03-11

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

  3. NEUTRONIC REACTOR MANIPULATING DEVICE

    DOEpatents

    Ohlinger, L.A.

    1962-08-01

    A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

  4. Nuclear reactor

    DOEpatents

    Thomson, Wallace B.

    2004-03-16

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  5. Documented Safety Analysis Addendum for the Neutron Radiography Reactor Facility Core Conversion

    SciTech Connect

    Boyd D. Christensen

    2009-05-01

    The Neutron Radiography Reactor Facility (NRAD) is a Training, Research, Isotope Production, General Atomics (TRIGA) reactor which was installed in the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) in the mid 1970s. The facility provides researchers the capability to examine both irradiated and non-irradiated materials in support of reactor fuel and components programs through non-destructive neutron radiography examination. The facility has been used in the past as one facet of a suite of reactor fuels and component examination facilities available to researchers at the INL and throughout the DOE complex. The facility has also served various commercial research activities in addition to the DOE research and development support. The reactor was initially constructed using Fuel Lifetime Improvement Program (FLIP)- type highly enriched uranium (HEU) fuel obtained from the dismantled Puerto Rico Nuclear Center (PRNC) reactor. In accordance with international non-proliferation agreements, the NRAD core will be converted to a low enriched uranium (LEU) fuel and will continue to utilize the PRNC control rods, control rod drives, startup source, and instrument console as was previously used with the HEU core. The existing NRAD Safety Analysis Report (SAR) was created and maintained in the preferred format of the day, combining sections of both DOE-STD-3009 and Nuclear Regulatory Commission Regulatory Guide 1.70. An addendum was developed to cover the refueling and reactor operation with the LEU core. This addendum follows the existing SAR format combining required formats from both the DOE and NRC. This paper discusses the project to successfully write a compliant and approved addendum to the existing safety basis documents.

  6. Gaseous fuel nuclear reactor research

    NASA Technical Reports Server (NTRS)

    Schwenk, F. C.; Thom, K.

    1975-01-01

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

  7. Nuclear Reactor Engineering Analysis Laboratory

    SciTech Connect

    Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

    1998-12-31

    The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels.

  8. R&D around a photoneutralizer-based NBI system (Siphore) in view of a DEMO Tokamak steady state fusion reactor

    NASA Astrophysics Data System (ADS)

    Simonin, A.; Achard, Jocelyn; Achkasov, K.; Bechu, S.; Baudouin, C.; Baulaigue, O.; Blondel, C.; Boeuf, J. P.; Bresteau, D.; Cartry, G.; Chaibi, W.; Drag, C.; de Esch, H. P. L.; Fiorucci, D.; Fubiani, G.; Furno, I.; Futtersack, R.; Garibaldi, P.; Gicquel, A.; Grand, C.; Guittienne, Ph.; Hagelaar, G.; Howling, A.; Jacquier, R.; Kirkpatrick, M. J.; Lemoine, D.; Lepetit, B.; Minea, T.; Odic, E.; Revel, A.; Soliman, B. A.; Teste, P.

    2015-11-01

    Since the signature of the ITER treaty in 2006, a new research programme targeting the emergence of a new generation of neutral beam (NB) system for the future fusion reactor (DEMO Tokamak) has been underway between several laboratories in Europe. The specifications required to operate a NB system on DEMO are very demanding: the system has to provide plasma heating, current drive and plasma control at a very high level of power (up to 150 MW) and energy (1 or 2 MeV), including high performances in term of wall-plug efficiency (η  >  60%), high availability and reliability. To this aim, a novel NB concept based on the photodetachment of the energetic negative ion beam is under study. The keystone of this new concept is the achievement of a photoneutralizer where a high power photon flux (~3 MW) generated within a Fabry-Perot cavity will overlap, cross and partially photodetach the intense negative ion beam accelerated at high energy (1 or 2 MeV). The aspect ratio of the beam-line (source, accelerator, etc) is specifically designed to maximize the overlap of the photon beam with the ion beam. It is shown that such a photoneutralized based NB system would have the capability to provide several tens of MW of D0 per beam line with a wall-plug efficiency higher than 60%. A feasibility study of the concept has been launched between different laboratories to address the different physics aspects, i.e. negative ion source, plasma modelling, ion accelerator simulation, photoneutralization and high voltage holding under vacuum. The paper describes the present status of the project and the main achievements of the developments in laboratories.

  9. Mass measurements on stable nuclides in the rare-earth region with the Penning-trap mass spectrometer TRIGA-TRAP

    NASA Astrophysics Data System (ADS)

    Ketelaer, J.; Audi, G.; Beyer, T.; Blaum, K.; Block, M.; Cakirli, R. B.; Casten, R. F.; Droese, C.; Dworschak, M.; Eberhardt, K.; Eibach, M.; Herfurth, F.; Minaya Ramirez, E.; Nagy, Sz.; Neidherr, D.; Nörtershäuser, W.; Smorra, C.; Wang, M.

    2011-07-01

    The masses of 15 stable nuclides in the rare-earth region have been measured with the Penning-trap mass spectrometer TRIGA-TRAP. This is the first series of absolute mass measurements linking these nuclides to the atomic-mass standard C12. Previously, nuclear reaction studies almost exclusively determined the literature values of these masses in the Atomic-Mass Evaluation. The TRIGA-TRAP results show deviations on the order of 3-4 standard deviations from the latest published values of the Atomic-Mass Evaluation 2003 for some cases. However, the binding-energy differences that are important for nuclear structure studies have been confirmed and improved. The new masses are discussed in the context of valence proton-neutron interactions using double differences of binding energies, δVpn(Z,N).

  10. Mass measurements on stable nuclides in the rare-earth region with the Penning-trap mass spectrometer TRIGA-TRAP

    SciTech Connect

    Ketelaer, J.; Audi, G.; Beyer, T.; Blaum, K.; Block, M.; Dworschak, M.; Herfurth, F.; Cakirli, R. B.; Casten, R. F.; Droese, C.; Eberhardt, K.; Eibach, M.; Smorra, C.; Minaya Ramirez, E.; Nagy, Sz.; Neidherr, D.; Noertershaeuser, W.; Wang, M.

    2011-07-15

    The masses of 15 stable nuclides in the rare-earth region have been measured with the Penning-trap mass spectrometer TRIGA-TRAP. This is the first series of absolute mass measurements linking these nuclides to the atomic-mass standard {sup 12}C. Previously, nuclear reaction studies almost exclusively determined the literature values of these masses in the Atomic-Mass Evaluation. The TRIGA-TRAP results show deviations on the order of 3-4 standard deviations from the latest published values of the Atomic-Mass Evaluation 2003 for some cases. However, the binding-energy differences that are important for nuclear structure studies have been confirmed and improved. The new masses are discussed in the context of valence proton-neutron interactions using double differences of binding energies, {delta}V{sub pn}(Z,N).

  11. Thermionic Reactor Design Studies

    SciTech Connect

    Schock, Alfred

    1994-06-01

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full

  12. NASA Reactor Facility Hazards Summary. Volume 1

    NASA Technical Reports Server (NTRS)

    1959-01-01

    The Lewis Research Center of the National Aeronautics and Space Administration proposes to build a nuclear research reactor which will be located in the Plum Brook Ordnance Works near Sandusky, Ohio. The purpose of this report is to inform the Advisory Committee on Reactor Safeguards of the U. S. Atomic Energy Commission in regard to the design Lq of the reactor facility, the characteristics of the site, and the hazards of operation at this location. The purpose of this research reactor is to make pumped loop studies of aircraft reactor fuel elements and other reactor components, radiation effects studies on aircraft reactor materials and equipment, shielding studies, and nuclear and solid state physics experiments. The reactor is light water cooled and moderated of the MTR-type with a primary beryllium reflector and a secondary water reflector. The core initially will be a 3 by 9 array of MTR-type fuel elements and is designed for operation up to a power of 60 megawatts. The reactor facility is described in general terms. This is followed by a discussion of the nuclear characteristics and performance of the reactor. Then details of the reactor control system are discussed. A summary of the site characteristics is then presented followed by a discussion of the larger type of experiments which may eventually be operated in this facility. The considerations for normal operation are concluded with a proposed method of handling fuel elements and radioactive wastes. The potential hazards involved with failures or malfunctions of this facility are considered in some detail. These are examined first from the standpoint of preventing them or minimizing their effects and second from the standpoint of what effect they might have on the reactor facility staff and the surrounding population. The most essential feature of the design for location at the proposed site is containment of the maximum credible accident.

  13. Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems

    NASA Astrophysics Data System (ADS)

    Alipchenkov, V. M.; Anfimov, A. M.; Afremov, D. A.; Gorbunov, V. S.; Zeigarnik, Yu. A.; Kudryavtsev, A. V.; Osipov, S. L.; Mosunova, N. A.; Strizhov, V. F.; Usov, E. V.

    2016-02-01

    The conceptual fundamentals of the development of the new-generation system thermal-hydraulic computational HYDRA-IBRAE/LM code are presented. The code is intended to simulate the thermalhydraulic processes that take place in the loops and the heat-exchange equipment of liquid-metal cooled fast reactor systems under normal operation and anticipated operational occurrences and during accidents. The paper provides a brief overview of Russian and foreign system thermal-hydraulic codes for modeling liquid-metal coolants and gives grounds for the necessity of development of a new-generation HYDRA-IBRAE/LM code. Considering the specific engineering features of the nuclear power plants (NPPs) equipped with the BN-1200 and the BREST-OD-300 reactors, the processes and the phenomena are singled out that require a detailed analysis and development of the models to be correctly described by the system thermal-hydraulic code in question. Information on the functionality of the computational code is provided, viz., the thermalhydraulic two-phase model, the properties of the sodium and the lead coolants, the closing equations for simulation of the heat-mass exchange processes, the models to describe the processes that take place during the steam-generator tube rupture, etc. The article gives a brief overview of the usability of the computational code, including a description of the support documentation and the supply package, as well as possibilities of taking advantages of the modern computer technologies, such as parallel computations. The paper shows the current state of verification and validation of the computational code; it also presents information on the principles of constructing of and populating the verification matrices for the BREST-OD-300 and the BN-1200 reactor systems. The prospects are outlined for further development of the HYDRA-IBRAE/LM code, introduction of new models into it, and enhancement of its usability. It is shown that the program of development and

  14. Autonomous Control of Space Nuclear Reactors

    NASA Technical Reports Server (NTRS)

    Merk, John

    2013-01-01

    Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation. Long-duration surface missions necessitate reliable autonomous operation, and manned missions impose added requirements for failsafe reactor protection. There is a need for an advanced instrumentation and control system for space-nuclear reactors that addresses both aspects of autonomous operation and safety. The Reactor Instrumentation and Control System (RICS) consists of two functionally independent systems: the Reactor Protection System (RPS) and the Supervision and Control System (SCS). Through these two systems, the RICS both supervises and controls a nuclear reactor during normal operational states, as well as monitors the operation of the reactor and, upon sensing a system anomaly, automatically takes the appropriate actions to prevent an unsafe or potentially unsafe condition from occurring. The RPS encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device output terminals. The SCS contains a comprehensive data acquisition system to measure continuously different groups of variables consisting of primary measurement elements, transmitters, or conditioning modules. These reactor control variables can be categorized into two groups: those directly related to the behavior of the core (known as nuclear variables) and those related to secondary systems (known as process variables). Reliable closed-loop reactor control is achieved by processing the acquired variables and actuating the appropriate device drivers to maintain the reactor in a safe operating state. The SCS must prevent a deviation from the reactor nominal conditions by managing limitation functions in order to avoid RPS actions. The RICS has four identical redundancies that comply with physical separation, electrical isolation, and functional independence. This architecture complies with the

  15. Wavelet-Based Analytical Algorithm for Solving Steady-State Concentration in Immobilized Glucose Isomerase of Packed-Bed Reactor Model.

    PubMed

    Selvi, M Salai Mathi; Hariharan, G

    2016-08-01

    Wavelet method is a recently developed tool in applied mathematics. The mathematical model of the steady-state immobilized enzyme electrodes is discussed. This theoretical model is based on one-dimensional heat conduction equations containing a non-linear term related to Michaelis-Menten kinetics. An efficient Chebyshev wavelet-based technique is applied to solve the non-linear diffusion equation for the steady-state condition. A simple expression of the substrate concentration is obtained as a function of the Thiele modulus [Formula: see text] and [Formula: see text](kinetic parameter). The wavelet results are compared with the numerical and HPM solutions and found to be in good agreement. PMID:27161606

  16. Dosimetric feasibility study for an extracorporeal BNCT application on liver metastases at the TRIGA Mainz.

    PubMed

    Blaickner, M; Kratz, J V; Minouchehr, S; Otto, G; Schmidberger, H; Schütz, C; Vogtländer, L; Wortmann, B; Hampel, G

    2012-01-01

    This study investigates the dosimetric feasibility of Boron Neutron Capture Therapy (BNCT) of explanted livers in the thermal column of the research reactor in Mainz. The Monte Carlo code MCNP5 is used to calculate the biologically weighted dose for different ratios of the (10)B-concentration in tumour to normal liver tissue. The simulation results show that dosimetric goals are only partially met. To guarantee effective BNCT treatment the organ has to be better shielded from all gamma radiation. PMID:21872481

  17. [The Chinese nuclear test and 'atoms for peace' as a measure for preventing nuclear armament of Japan: the nuclear non-proliferation policy of the United States and the introduction of light water reactors into Japan, 1964-1968].

    PubMed

    Yamazaki, Masakatsu

    2014-07-01

    Japan and the United States signed in 1968 a new atomic energy agreement through which US light-water nuclear reactors, including those of the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company, were to be introduced into Japan. This paper studies the history of negotiations for the 1968 agreement using documents declassified in the 1990s in the US and Japan. After the success of the Chinese nuclear test in October 1964, the United States became seriously concerned about nuclear armament of other countries in Asia including Japan. Expecting that Japan would not have its own nuclear weapons, the US offered to help the country to demonstrate its superiority in some fields of science including peaceful nuclear energy to counter the psychological effect of the Chinese nuclear armament. Driven by his own political agenda, the newly appointed Prime Minister Eisaku Sato responded to the US expectation favorably. When he met in January 1965 with President Johnson, Sato made it clear that Japan would not pursue nuclear weapons. Although the US continued its support after this visit, it nevertheless gave priority to the control of nuclear technology in Japan through the bilateral peaceful nuclear agreement. This paper argues that the 1968 agreement implicitly meant a strategic measure to prevent Japan from going nuclear and also a tactic to persuade Japan to join the Nuclear Non -Proliferation Treaty. PMID:25296517

  18. Self-actuating reactor shutdown system

    DOEpatents

    Barrus, Donald M.; Brummond, Willian A; Peterson, Leslie F.

    1988-01-01

    A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

  19. Reactivity control assembly for nuclear reactor. [LMFBR

    DOEpatents

    Bollinger, L.R.

    1982-03-17

    This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

  20. Research reactor de-fueling and fuel shipment

    SciTech Connect

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

    1998-08-01

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