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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. Arkansas Tech University TRIGA nuclear reactor

    SciTech Connect

    Sankoorikal, J.; Culp, R.; Hamm, J.; Elliott, D.; Hodgson, L.; Apple, S.

    1990-07-01

    This paper describes the TRIGA nuclear reactor (ATUTR) proposed for construction on the campus of Arkansas Tech University in Russellville, Arkansas. The reactor will be part of the Center for Energy Studies located at Arkansas Tech University. The reactor has a steady state power level of 250 kW and can be pulsed with a maximum reactivity insertion of $2.0. Experience gained in dismantling and transporting some of the components from Michigan State University, and the storage of these components will be presented. The reactor will be used for education, training, and research. (author)

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

    SciTech Connect

    Anellis, L.G.; Johnson, A.G.; Higginbotham, J.F.

    1988-07-01

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

  4. Analysis of the TRIGA Reactor Pool Water

    DTIC Science & Technology

    1993-08-01

    AD-A270 956 L11L1I~I1 11 11 :1Ji ili! August 1993 AFRRI 93-5 TECHNICAL REPORT Analysis of the TRIGA Reactor Pool Water L OCT 1 93 John Dickson Robert...COVER~ED I August 1993 Technical Report 4 TITLE AND SUBTITLE S.FNDN NUMBERS Analysis of the TRIGA Reactor Pool Water PE: NWED QAXM 6, AUTHOR(S) Dickson...AVAILABIIY STATEMENT 1 2b. DISTRIBUTION CODE Approved for public release; distribution unlimited. 13. ABSTRACT tMaxtm -um 200 words ) 14. SUBJECTTERMS 1S

  5. Preparation and planning for the replacement of the Oregon State University TRIGA reactor rotary specimen rack assembly

    SciTech Connect

    Anderson, T.V.; Dodd, B.; Johnson, A.G.; Carpenter, W.T.

    1984-07-01

    Recently there have been a number of indications that the rotating rack may be approaching the end of its useful life. In order to benefit from the experience of other reactors who have removed and replaced their rotating racks, General Atomic (GA) was contacted and previous TRIGA Conference proceedings were scanned. It was determined that a number of facilities, had experienced difficulties with their lazy susans and eventually had to replace them. However, most of the written descriptions of this project were not sufficiently detailed to be of great use. The purpose of this paper is to identify some of the more important questions related to the replacement of our rotating rack assembly and OSU's currently proposed solutions, with a view towards soliciting ideas from other members of the TRIGA reactor community.

  6. 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

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

  7. Applications of the Dow TRIGA research reactor

    SciTech Connect

    Kocher, C.W.; Quinn, T.J.; Krueger, D.A.

    1982-07-01

    The Dow TRIGA Mark I reactor is a one-hundred kilowatt nuclear reactor installed by General Atomics using the Torrey Pines reactor console, seventy-five used stainless-steel clad fuel elements and one new aluminium clad fuel element. The reactor is equipped with a forty-position rotating Lazy Susan in the reflector, a pneumatic transfer system with its terminal in the F-ring of the core, and a central thimble which can be used for irradiation of samples in the center of the core or which can be emptied of the shielding water to produce a beam of neutrons and gamma rays in the area atop the pool. Samples can also be irradiated in or near the core. There is no provision for pulsing this TRIGA reactor. The neutron activation analysis program uses the Dow TRIGA reactor as a source of thermal neutrons and a Kaman A711 generator as a source of 14-MeV neutrons. The associated counting equipment includes one Gel(Li) detector and two Nal(Tl) detectors, each using a 100-position sample changer and all interfaced to a Tracor-Northern TN-11 data acquisition and computing system, one Ge(Li) detector and its TN-11 system for the pneumatic transfer system and the beam tube experiments, and two NaKTl)detectors with a TN-4000 system used with the Kaman neutron generator. The activation analysis program gets samples from all parts of the manufacturing and research efforts at Dow: raw materials, intermediates, products, effluents, research samples, samples from customers who use Dow products, and environmental samples. This presentation is devoted to the progress made in the past year on the pneumatic transfer system and the renewed work on prompt gamma-ray spectroscopy including the extensive process of method validation.

  8. Archaeometry using the Cornell TRIGA reactor

    SciTech Connect

    Hossain, T.Z.; Bentley, R.A.; Clark, D.D.; Keyser, P.T.; Sadlikova, K.; Silverman, A.; Whitehead, J.K.

    1994-12-31

    There are several archaeological studies in progress at Cornell, and the various excavations include Etruscan (Siena, Italy), Neolithic (Halai, Greece), and late Classical (Sula Valley, northwest Honduras) Honduran sites. As part of these investigations, samples from these sites are analyzed using the Cornell TRIGA reactor. Cornell TRIGA is a 500-kW reactor well suited for high-sensitivity neutron activation analysis (NAA) and prompt gamma NAA (PGNAA). The NAA facilities include several 20 to 30% efficient high-purity germanium detectors coupled to several personal computer-based analyzers. A PGNAA facility using a cold neutron beam is currently under development. The Cornell cold neutron beam is designed for a temperature of {approximately}30 K and a flux of {approximately}1 x 10{sup 7} n /cm{sup 2}{center_dot}s thermal equivalent. The guided beam is extracted and taken 13 m from the TRIGA core into an adjoining laboratory area, where the PGNAA irradiation and counting station will be set up.

  9. 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

    ... the licensee to operate the AFFRI TRIGA Reactor up to a steady-state thermal power of 1.1 MW for an additional 20 years from the date of issuance. DATES: Submit comments by December 17, 2012. Requests for a... authorize the licensee to operate the AFFRI TRIGA Reactor up to a steady-state thermal power of 1.1 MW...

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

    SciTech Connect

    ZEFRAN, BOJAN

    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.

  11. Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

    NASA Astrophysics Data System (ADS)

    Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Unruh, Troy C.; McGregor, Douglas S.; Roberts, Jeremy A.

    2017-08-01

    Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

  12. Upgrade of the Dow TRIGA research reactor

    SciTech Connect

    Kocher, C.W.

    1991-11-01

    Useful operation of the Dow TRIGA{sup a} research reactor over a period of >20 years has led to a commitment to upgrades enabling another two decades of use with increased capabilities. The reactor utilization program and the upgrades are described in this paper. These included requesting a 20-yr license instead of the 10-yr license, which had been used previously; changing the license to allow operation at power levels of up to 300 kW, which provided improved analytical sensitivity; adding fuel elements to the core, which allowed better performance at the higher power levels; renovating the laboratories, which included consolidating the radioactive materials handling areas and improving the sample preparation areas; installing new shielding, detectors, computers, and sample-handling robots for greater productivity and sensitivity; replacing the 1967-1974 era control console and renovating the control rod drives to provide greater safety, reliability, and maintenance capabilities; and identifying, training, and licensing more senior reactor operators to allow the staff to continue operating and improving this system well past the turn of the century.

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

    NASA Astrophysics Data System (ADS)

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

    2017-02-01

    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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

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

    SciTech Connect

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

    2016-01-22

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

  19. An Analysis of Decommissioning Costs for the AFRRI TRIGA Reactor Facility

    DTIC Science & Technology

    1989-12-01

    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.

  20. 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.

  1. Automated power control system for reactor TRIGA PUSPATI

    NASA Astrophysics Data System (ADS)

    Ghazali, Anith Khairunnisa; Minhat, Mohd Sabri; Hassan, Mohd Khair

    2017-01-01

    Reactor TRIGA PUSPATI (RTP) Mark II type undergoes safe operation for more than 30 years and the only research reactor exists in Malaysia. The main safety feature of Instrumentation and Control (I&C) system design is such that any failure in the electronic, or its associated components, does not lead to an uncontrolled rate of reactivity. The existed controller using feedback approach to control the reactor power. This paper introduces proposed controllers such as Model Reference Adaptive Control (MRAC) and Proportional Integral Derivatives (PID) controller for the RTP simulation. In RTP, the most important considered parameter is the reactor power and act as nervous system. To design a controller for complex plant like RTP is quite difficult due to high cost and safety factors cause by the failure of the controller. Furthermore, to overcome these problems, a simulator can be used to replace functions the hardware and test could then be simulated using this simulator. In order to find the best controller, several controllers were proposed and the result will be analysed for study the performances of the controller. The output result will be used to find out the best RTP power controller using MATLAB/Simulink and gives result as close as the real RTP performances. Currently, the structures of RTP was design using MATLAB/Simulink tool that consist of fission chamber, controller, control rod position, height-to-worth of control rods and a RTP model. The controller will control the control rod position to make sure that the reactivity still under the limitation parameter. The results given from each controller will be analysed and validated through experiment data collected from RTP.

  2. 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.

  3. 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

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION University of Utah, University of Utah TRIGA Nuclear Reactor, Notice of Issuance of Renewed... University of Utah (UU, the licensee), which authorizes continued operation of the UU TRIGA Nuclear Reactor...

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

    PubMed

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

    2017-09-09

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

  5. 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.

  6. 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.

  7. 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).

  8. 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)

  9. 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.

  10. 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.

  11. 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)

  12. 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.

  13. 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.

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

    SciTech Connect

    Unlu, Kenan; Bauer, Thomas L.; Wehring, Bernard W.

    1992-07-01

    Recently, the UT-TRIGA research reactor was licensed and has become fully operational. This reactor, the first new US university reactor in 17 years, is the focus of a new reactor laboratory facility which is located on the Balcones Research Center at The University of Texas at Austin. The TRIGA Mark II reactor is licensed for 1.1 MW steady power operation, 3 dollar pulsing, and includes five beam ports. Various neutron beam-line projects have been assigned to each beam port. Neutron Depth Profiling (NDP) and the Texas Cold Neutron Source (TCNS) are close to completion and will be operational in the near future. The design of the NDP instrument has been completed, a target chamber has been built, and the thermal neutron collimator, detectors, data acquisition electronics, and data processing computers have been acquired. The target chamber accommodates wafers up to 12'' in diameter and provides remote positioning of these wafers. The design and construction of the TCNS has been completed. The TCNS consists of a moderator (mesitylene), a neon heat pipe, a cryogenic refrigerator, and neutron guide tubes. In addition, fission-fragment research (HIAWATHA), Neutron Capture Therapy, and Neutron Radiography are being pursued as projects for the other three beam ports. (author)

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

    SciTech Connect

    Aderhold, Howard C.

    1984-07-01

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

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

    PubMed

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

    2017-02-01

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

  17. 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.

  18. 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.

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

    PubMed

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

    2013-01-01

    This work presents an initial proposed design of a Prompt Gamma Activation Analysis (PGAA) facility to be installed at the TRIGA IPR-R1, a 60 years old research reactor of the Centre of Development of Nuclear Technology (CDTN) in Brazil. The basic characteristics of the facility and the results of the neutron flux are presented and discussed. The proposed design is based on a quasi vertical tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below the room's level where shall be located the rack containing the set sample/detector/shielding. The evaluation of the thermal and epithermal neutron flux in the sample position was done considering the experimental data obtained from a vertical neutron guide, already existent in the reactor, and the simulated model for the facility. The experimental determination of the neutron flux was obtained through the standard procedure of using Au monitors in different positions of the vertical tube. In order to validate both, this experiment and calculations of the simulated model, the flux was also determined in different positions in the core used for sample irradiation. The model of the system was developed using the Monte Carlo code MCNP5. The preliminary results suggest the possibility of obtaining a beam with minimum thermal flux of magnitude 10(6) cm(-2) s(-1), which confirm the technical feasibility of the installation of PGAA at the TRIGA IPR-R1 reactor. This beam would open new possibilities for enhancing the applications using the reactor.

  20. 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.

  1. Neutronics and thermal hydraulic analysis of TRIGA Mark II reactor using MCNPX and COOLOD-N2 computer code

    NASA Astrophysics Data System (ADS)

    Tiyapun, K.; Wetchagarun, S.

    2017-06-01

    The neutronic analysis of TRIGA Mark II reactor has been performed. A detailed model of the reactor core was conducted including standard fuel elements, fuel follower control rods, and irradiation devices. As the approach to safety nuclear design are based on determining the criticality (keff), reactivity worth, reactivity excess, hot rod power factor and power peaking of the reactor, the MCNPX code had been used to calculate the nuclear parameters for different core configuration designs. The thermal-hydraulic model has been developed using COOLOD-N2 for steady state, using the nuclear parameters and power distribution results from MCNPX calculation. The objective of the thermal-hydraulic model is to determine the thermal safety margin and to ensure that the fuel integrity is maintained during steady state as well as during abnormal condition at full power. The hot channel fuel centerline temperature, fuel surface temperature, cladding surface temperature, the departure from nucleate boiling (DNB) and DNB ratio were determined. The good agreement between experimental data and simulation concerning reactor criticality proves the reliability of the methodology of analysis from neutronic and thermal hydraulic perspective.

  2. 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

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Notice of License Termination for University of Illinois Advanced TRIGA Reactor, License No. R-115 The U.S. Nuclear Regulatory Commission (NRC) is noticing the termination of Facility Operating...

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

    SciTech Connect

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

    2016-01-22

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

  4. 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.

  5. 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

  6. Mixed core calculations for the TRIGA reactor of Mexico

    SciTech Connect

    Mazon-Ramirez, R.; Francois, J.L; Gallardo, L.F.

    1990-07-01

    It was decided to use FLIP fuel (70 % enrichment in U235) - against the world tendency to reduce enrichment- because we already had 46 fresh FLIP fuels available in the facility and almost non of the standard type. In order to better use the fuel we had at that time, we decided to find a load scheme that better suited that purpose, we could use 46 FLIP fuel elements no irradiated and about 120 irradiated standard fuel elements. We use EXTERMINATOR to analize several options. We studied the burnup of the reactor core consisting of the two types of fuel, the total numbers of fuel elements was kept constant 85 to compare directly the results. The numbers of inserted FLIP elements was adjusted trying to get realistic core excess reactivity (about 6 $) and trying not to complicate power shapes mixing the two types of fuel in one ring. We selected two of them that were analized in more detail. The first one we called 8th configuration used all 46 FLIP elements placed in ring C, D and E with 59 irradiated standard elements filling the other positions, and the 9th configuration using 26 FLIP fuel elements in rings C and D with irradiated standard fuel completing the load. In both cases the three burned control rods were sustituted with new FLIP follower control rods. After some rough calculations we decided to drop the 8th configuration because we were concerned with the shoutdown margin of that configuration. After we loaded the fuel and performed rod worth measurements we finded out that we were subestimating the worth in our calculations. (author)

  7. 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)

  8. 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.

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

    NASA Astrophysics Data System (ADS)

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

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

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. Assessment and mitigation of power quality problems for PUSPATI TRIGA Reactor (RTP)

    NASA Astrophysics Data System (ADS)

    Zakaria, Mohd Fazli; Ramachandaramurthy, Vigna K.

    2017-01-01

    An electrical power systems are exposed to different types of power quality disturbances. Investigation and monitoring of power quality are necessary to maintain accurate operation of sensitive equipment especially for nuclear installations. This paper will discuss the power quality problems observed at the electrical sources of PUSPATI TRIGA Reactor (RTP). Assessment of power quality requires the identification of any anomalous behavior on a power system, which adversely affects the normal operation of electrical or electronic equipment. A power quality assessment involves gathering data resources; analyzing the data (with reference to power quality standards) then, if problems exist, recommendation of mitigation techniques must be considered. Field power quality data is collected by power quality recorder and analyzed with reference to power quality standards. Normally the electrical power is supplied to the RTP via two sources in order to keep a good reliability where each of them is designed to carry the full load. The assessment of power quality during reactor operation was performed for both electrical sources. There were several disturbances such as voltage harmonics and flicker that exceeded the thresholds. To reduce these disturbances, mitigation techniques have been proposed, such as to install passive harmonic filters to reduce harmonic distortion, dynamic voltage restorer (DVR) to reduce voltage disturbances and isolate all sensitive and critical loads.

  20. 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

  1. 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.

  2. 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

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

    SciTech Connect

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

    1992-07-01

    Many TRIGA reactors have experienced operational difficulties with the rotating racks used for sample irradiation. Generally the rack gradually becomes more difficult to rotate until it finally seizes. The recommended action at that point is replacement of the entire facility at a significant cost. The purpose of this paper is to describe the symptoms leading to rack failure and to present the results of a refurbishment procedure that does not involve the use of solvents which create mixed chemical and radioactive hazardous waste. The primary reason for rack failure is the buildup of sludge produced through irradiation of lubrication oil. The refurbishment procedure involves using a commercially available degreasing solution which can be pumped into and out of the rack with the objective of removing this sludge. The solution used is sold under the trade name 'Simple Green'. No radioactive material was detected on smear or air samples taken of the work area during the reifurbishment activities and the rack rotates freely in both direction even after eighteen months of operation. The only disadvantage to performing this procedure has been the need to maintain a very aggressive contamination control program when unloading samples from the rack. A very fine particulate material attaches to the outside of tubes used to encapsulate samples. This material can produce contamination levels of 10,000 dpm/100 cm{sup 2} in the worst cases but will typically produce local hot spots on the order of 1000 dpm. (author)

  4. 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.

  5. 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.

  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. Copyright © 2015 Elsevier Ltd. All rights reserved.

  7. 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.

  8. 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

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

    SciTech Connect

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

    2016-07-12

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

  10. 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.

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

    SciTech Connect

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

    2016-01-22

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

  13. 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.

  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. 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.

  16. 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.

  17. Adaption of a PIN-diode detector as an online neutron monitor for the thermal column of the TRIGA research reactor.

    PubMed

    Schmitz, Tobias; Blaickner, Matthias

    2017-10-01

    A BNCT online neutron monitoring system was tested in a TRIGA reactor, using a silicon PIN-diode with a conversion foil. The setup was tested with different reactor powers at the hot and cold ends of the irradiation channel, using activation foils to compare with measured fluxes. The results demonstrate good reproducibility and show a linear correlation between signal of the PIN-diode and neutron flux at all positions, demonstrating this approach to be suitable for online monitoring of the neutron flux. Copyright © 2017 Elsevier Ltd. All rights reserved.

  18. 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.

  19. Benchmarking criticality analysis of TRIGA fuel storage racks.

    PubMed

    Robinson, Matthew Loren; DeBey, Timothy M; Higginbotham, Jack F

    2017-01-01

    A criticality analysis was benchmarked to sub-criticality measurements of the hexagonal fuel storage racks at the United States Geological Survey TRIGA MARK I reactor in Denver. These racks, which hold up to 19 fuel elements each, are arranged at 0.61m (2 feet) spacings around the outer edge of the reactor. A 3-dimensional model was created of the racks using MCNP5, and the model was verified experimentally by comparison to measured subcritical multiplication data collected in an approach to critical loading of two of the racks. The validated model was then used to show that in the extreme condition where the entire circumference of the pool was lined with racks loaded with used fuel the storage array is subcritical with a k value of about 0.71; well below the regulatory limit of 0.8. A model was also constructed of the rectangular 2×10 fuel storage array used in many other TRIGA reactors to validate the technique against the original TRIGA licensing sub-critical analysis performed in 1966. The fuel used in this study was standard 20% enriched (LEU) aluminum or stainless steel clad TRIGA fuel.

  20. 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.

  1. 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

  2. 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...

  3. Optimised k0-instrumental neutron activation method using the TRIGA MARK I IPR-R1 reactor at CDTN/CNEN, Belo Horizonte, Brazil

    NASA Astrophysics Data System (ADS)

    Menezes, M. Â. B. C.; Jaćimović, R.

    2006-08-01

    The Nuclear Technology Development Centre/Brazilian Commission for Nuclear Energy, CDTN/CNEN, is the only Brazilian Institution to apply the k0-standardisation method of instrumental neutron activation technique determining elements using its own nuclear reactor, TRIGA MARK I IPR-R1. After changes in the reactor core configuration, the reactor neutron flux distribution in typical irradiation channels had to be updated, as well as the parameters f and α, needed to apply the k0-method of neutron activation analysis. The neutron flux distribution in the rotary rack was evaluated through the specific count rate of 198Au and the parameters f and α, were determined in five selected channels applying the "Cd-ratio for multi-monitor" method, using a set of Al-(0.1%)Au and Zr (99.8%) monitors. Several reference materials were analysed, indicating the effectiveness of the improved method.

  4. 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.

  5. 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.

  6. 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)

  7. 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.

  8. 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.

  9. 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.

  10. 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

  11. 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).

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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

  18. Modification of the rotary specimen rack (Lazy Susan) at the U.S. Geological Survey TRIGA Reactor (GSTR) to permit pneumatic unloading of samples

    SciTech Connect

    Heifer, P.G.; Rusling, D.H.

    1986-07-01

    With a continuing sample load of about 18,000 samples per year, the U.S.G.S. TRIGA facility has installed a modified rotary specimen rack that allows samples to be unloaded from the rack using a closed loop blower system. The radiation exposure to operating personnel is lowered due to remote operation. Time required to unload large numbers of samples is reduced dramatically, and the sample capacity of the lazy susan can be doubled due to design changes of the. irradiation capsule allowed by the pneumatic unloading system. The safety, versatility, time and costs are improved by the addition of the pneumatic unloading lazy susan. (author)

  19. 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.

  20. 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.

  1. 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.

  2. Advanced BWR stability monitoring tests with a hybrid reactor facility

    SciTech Connect

    He, Weidong; Huang, Zhengyu; Edwards, Robert M.

    2002-07-01

    A stability monitor for a boiling reactor is implemented and evaluated with the Penn State Hybrid Reactor System. The stability monitor is based on an extended Kalman filter which employs a reduced-order BWR reactor model. The filter uses measured power signal and estimates the void reactivity feedback gain and the decay ratio. The hybrid reactor system is a system combining a simulation module of BWR thermal hydraulics and the Penn State TRIGA reactor. A description of the hybrid system is also presented. (authors)

  3. 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.

  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. Temperature feedback of TRIGA MARK-II fuel

    SciTech Connect

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

    2016-01-22

    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.

  6. 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.

  7. 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

  8. 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.

  9. Decommissioning of the TRIGA mark II and III and radioactive waste management

    SciTech Connect

    Doo Seong Hwang; Yoon Ji Lee; Gyeong Hwan Chung; Jei Kwon Moon

    2013-07-01

    KAERI has carried out decommissioning projects for two research reactors (KRR-1 and 2). The decommissioning project of KRR-1 (TRIGA Mark II) and 2 (TRIGA Mark III) was launched in 1997 with a total budget of 23.25 million US dollars. KRR-2 and all auxiliary facilities were already decommissioned, and KRR-1 is being decommissioned now. Much more dismantled waste is generated than in any other operations of nuclear facilities. Thus, the waste needs to be reduced and stabilized through decontamination or treatment before disposal. This paper introduces the current status of the decommissioning projects and describes the volume reduction and conditioning of decommissioning waste for final disposal. (authors)

  10. High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

    SciTech Connect

    Perez, Emmanuel; Keiser, Jr., Dennis D.; Forsmann, Bryan; Janney, Dawn E.; Henley, Jody; Woolstenhulme, Eric C.

    2016-02-01

    High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. Refurbishment of the Oregon State University rotating rack

    SciTech Connect

    Higginbotham, J.F. )

    1991-01-01

    TRIGA reactors have experienced operational difficulties with the rotating racks used for sample irradiation. The most common problem occurs when the rack seizes, and the corrective action taken is replacement of the rack assembly. This paper describes the symptoms leading to rack failure and a refurbishment procedure to correct the problem without replacing the rack at the Oregon State University TRIGA Reactor (OSTR) Facility. This procedure was accomplished with extraordinary results from an operational and a radiation protection standpoint. The refurbishment has extended the useful life of this reactor facility with minimal financial impact. Given the declining number of university-based research reactors, it is in the nation's best interest to maintain the currently operating research reactor facilities, and the described procedure can aid in achieving that goal.

  17. Modeling of γ field around irradiated TRIGA fuel elements by R2S method

    NASA Astrophysics Data System (ADS)

    Klemen, Ambrožič; Luka, Snoj

    2017-09-01

    The JSI TRIGA reactor has several irradiation facilities with well characterized neutron fields. The characterization was performed by measurements and by utilizing Monte Carlo particle transport computational methods. Because of this, JSI TRIGA has become a reference center for neutron irradiation of detectors for ATLAS experiment (CERN). Thorough γ characterization of the reactor is however yet to be performed. Current Monte Carlo particle transport code only account for the prompt generation of neutron induced γ rays, which have been characterized, but are neglecting the time dependent delayed part, which may in some cases amount to more then 30% of total γ flux in an operation reactor, and is the only source of γ-rays after reactor shutdown. Several common approaches of modeling delayed -rays , namely D1S and R2S exist. In this paper an in-house developed R2S method code is described, coupling a Monte Carlo particle transport code MCNP6 and neutron activation code FISPACT-II, with intermediate steps performed by custom Python scripts. An example of its capabilities is presented in terms of evaluation of utilization of JSI TRIGA nuclear fuel as a viable γ-ray source. In the model, fresh nuclear fuel is considered and a silicon pipe sample is modeled in. Fuel activities, dose and kerma rates on the sample, as well as emitted γ-ray spectra and isotopic contribution to the contact dose are calculated and presented.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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

  4. 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.

  5. 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...

  6. 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.

  7. 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

    ... Rulemaking, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Rockville, MD 20852... operation of the WSU Modified TRIGA Nuclear Reactor, located in Pullman, Whitman County, Washington. The WSU Modified TRIGA Nuclear Reactor is a pool-type, natural convection, light-water cooled, and shielded...

  8. Solid State Track Recorder applications in United States nuclear reactor energy programs

    SciTech Connect

    Gold, R.; Ruddy, F.H.; Roberts, J.H.

    1980-01-01

    The domain of Solid State Track Recorder (SSTR) applications in United States nuclear reactor energy programs extends from the harsh high temperature environments found in high power reactor cores to very low flux environments arising in out-of-core locations, critical assemblies, or away from reactors (AFR) experiments. The neutron energy region arising in these applications is very broad, covering more than eight orders of magnitude from thermal up to fusion energies. The range of neutron flux/fluence intensity is even greater, extending over more than thirteen orders of magnitude. As a consequence, use of a variety of SSTR is entailed in US nuclear reactor energy programs.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. Steady State Analysis of Small Molten Salt Reactor

    NASA Astrophysics Data System (ADS)

    Yamamoto, Takahisa; Mitachi, Koshi; Suzuki, Takashi

    The Molten Salt Reactor (MSR) is a thermal neutron reactor with graphite moderation and operates on the thorium-uranium fuel cycle. The feature of the MSR is that fuel salt flows inside the reactor during the nuclear fission reaction. In the previous study, the authors developed numerical model with which to simulate the effects of fuel salt flow on the reactor characteristics. In this study, we apply the model to the steady-state analysis of a small MSR system and estimate the effects of fuel flow. The model consists of two-group neutron diffusion equations for fast and thermal neutron fluxes, transport equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and the graphite moderator. The following results are obtained: (1) in the rated operation condition, the peaks of the neutron fluxes slightly move toward the bottom from the center of the reactor and the delayed neutron precursors are significantly carried by the fuel salt flow, and (2) the extension of residence time in the external-loop system and the rise of the fuel inflow temperature show weak negative reactivity effects, which decrease the neutron multiplication factor of the small MSR system.

  14. 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.

  15. 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.

  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. 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

  18. 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.

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

    SciTech Connect

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

    2016-01-22

    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 UZrH{sub 1.6}, 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{sup ®} 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.

  20. 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.

  1. 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... potential severe reactor accidents for the Peach Bottom Atomic Power Station and the Surry Power Station....

  2. A non-inductively driven steady state tokamak reactor

    SciTech Connect

    Fenstermacher, M.E.; Devoto, R.S.; Bulmer, R.H.; Lee, J.D.; Miller, J.R.; Schultz, J.

    1988-09-20

    The physics and engineering guidelines for the ITER device are shown to lead to viable and attractive operating points for a steady state tokamak power reactor. Non-inductive current drive is provided in steady state by high energy neutral beam injection in the plasma core, lower hybrid slow waves in the outer regions of the plasma and bootstrap current. Plasma gain Q (/equivalent to/fusion power/input power) in excess of 20 and average neutron wall loading, approx. 2.0 MW/m/sup 2/ are predicted in a device with major radius, R/sub 0/ = 7.5 m and minor radius, a = 2.8 m. 15 refs., 3 figs., 3 tabs.

  3. 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.

  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. 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.

  7. 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.

  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. Recent accomplishments in neutron beam projects at the University of Texas Research Reactor

    SciTech Connect

    Uenlue, K.; Wehring, B.W.

    1994-12-31

    The design of a cold neutron source facility at the University of Texas TRIGA research reactor is described. The UT-TRIGA has 5 neutron beam ports. Because of the different characteristics of the ports, various research projects are being pursued. Among these projects, The Texas cold neutron source and neutron depth profiling are operational; neutron focusing, prompt gamma activation analysis, and neutron capture therapy research are progressing.

  10. 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)

  11. Neutron flux and spectrum characterization in the University of Illinois TRIGA

    SciTech Connect

    Peach, R.O.; Williams, J.G.; Ougouag, A.M.

    1988-07-01

    The principal irradiation facilities at the University of Illinois TRIGA are a central thimble, a pneumatic 'rabbit' tube in the G-hexagon, a forty- position Lazy Susan facility in the reflector, a through-port passing tangentially through the reflector, and a recently installed rotating tube in a three-element cluster position in the F- and G-hexagons. These facilities are used primarily for thermal and epithermal neutron activation, and for radiation damage studies in semiconductor and other materials. Thermal and epithermal neutron fluence rates have recently been measured by means of radiometric monitors in all the core and reflector facilities at a variety of reactor power levels. In addition, the fast neutron fluences have been characterized by means of threshold and non-threshold activation reaction rates measured in those positions used for radiation damage studies. Unfolding of neutron spectra from the measured reaction rates was done by means of the code LSL-M2, which was made available in 1986 from the Radiation Shielding Information Center. This code uses the generalized least squares method and incorporates a full propagation of uncertainties due to errors in measured reaction rates, cross section data, and input neutron spectra. The input spectra, needed in LSL-M2, were taken from the General Atomic report GA4361 which includes 24-group spectra, calculated by the code GAZE, for the center of the core, the F-hexagon and the Lazy Susan. These positions, and some of the material and geometrical specifications, do not correspond exactly with the actual ones at the University of Illinois TRIGA, but the differences in the neutron spectra were found to be small. The unfolded neutron spectra may be used to obtain fast neutron spectrum parameters needed in radiation damage studies and hardness testing. For example, the 1-MeV equivalent fluence rates for displacement damage in silicon, calculated according to the ASTM Standard E722, are reported. (author)

  12. 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)

  13. 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.

  14. Description of the Steady-State Operation of a Biochemical Reactor Using a Diffusion Model

    NASA Astrophysics Data System (ADS)

    Moshinskii, A. I.

    2017-07-01

    Consideration has been given to nonlinear boundary-value problems of steady-state operation of a biochemical reactor. Nontrivial solutions of these problems have been found using equations of a diffusion model with homogeneous boundary conditions. A study has been made of the operating regimes of the reactor under different conditions of biomass transfer in it.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. Thin foil Faraday collector under burning plasma reactor like conditions

    NASA Astrophysics Data System (ADS)

    Aakhus-Witt, Andy; Hawbaker, James; Sayers, Jack; Cecil, Ed; Debay, Tim

    2002-04-01

    We are investigating thin foil Faraday collectors as a diagnostic for lost alpha particles from high yield d-t tokamak fusion plasmas. One advantage of this concept is the predicted ability to operate under fairly high neutron/gamma radiation backgrounds. We have tested this prediction by measuring the current from a thin (2.5 n) Ni foil placed in the core of the TRIGA fission reactor at the Denver Federal Center. At a maximum steady state power of 950 kW (1E13 n/cm2/s) a current of 1.2 nA/cm2 was measured. This measured current at the maximum power is somewhat larger than the calculated current (0.5 nA/cm2). These results, future planned measurements, and the application of this diagnostic to proposed burning plasma (BPX) experiments will be discussed.

  20. Vibro-acoustic Imaging at the Breazeale Reactor

    SciTech Connect

    Smith, James Arthur; Jewell, James Keith; Lee, James Edwin

    2016-09-01

    The INL is developing Vibro-acoustic imaging technology to characterize microstructure in fuels and materials in spent fuel pools and within reactor vessels. A vibro-acoustic development laboratory has been established at the INL. The progress in developing the vibro-acoustic technology at the INL is the focus of this report. A successful technology demonstration was performed in a working TRIGA research reactor. Vibro-acoustic imaging was performed in the reactor pool of the Breazeale reactor in late September of 2015. A confocal transducer driven at a nominal 3 MHz was used to collect the 60 kHz differential beat frequency induced in a spent TRIGA fuel rod and empty gamma tube located in the main reactor water pool. Data was collected and analyzed with the INLDAS data acquisition software using a short time Fourier transform.

  1. 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

  2. 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.

  3. Determination of the steady-state behavior of immobilized. beta. -galactosidase utilizing an integral reactor scheme

    SciTech Connect

    Scott, T.C.; Hill, C.G. Jr.; Amundson, C.H.

    1985-01-01

    Analysis of the steady-state behavior of immobilized ..beta..-galactosidase by integral reactor techniques has yielded a model which allows one to predict reactor performance under normal operating conditions. Values of the mechanistic rate constants for enzymatic hydrolysis of lactose were determined as a function of temperature by fitting the reactor model to experimental lactose conversion profiles. Use of this model along with the activity decay characteristics of the immobilized enzyme which have been presented in a previous publication could prove to be a useful tool in determining appropriate operating strategies for industrial applications of the immobilized enzyme catalyst. 18 refs., 5 figs., 6 tabs.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. Use of research reactors in multidisciplinary education at Cornell University

    SciTech Connect

    Clark, D.D. )

    1992-01-01

    Multidisciplinary aspects of nuclear science and technology form a large part of the research and teaching activities of the Nuclear Science and Engineering (NS and E) Program at Cornell, and the two reactors housed in Ward Laboratory - a 500-kW TRIGA and a 100-W critical facility (zero-power reactor (ZPR))- play a central role in those activities. Several primarily educational and multidisciplinary features of the NS and E program are described in this paper.

  9. 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.

  10. Effect of sludge discharge positions on steady-state aerobic granules in sequencing batch reactor (SBR).

    PubMed

    Liu, Lin; Gao, Da-Wen; Liang, Hong

    2012-01-01

    We have investigated the effect of sludge discharge location on the steady-state aerobic granules in sequencing batch reactors (SBRs). Two SBRs were operated concurrently with the same sludge retention time using sludge discharge ports at: (a) the reactor bottom in R1; and (b) the reactor middle-lower level in R2. Results indicate that both reactors could maintain sludge granulation and stable operation, but the two different sludge discharge methods resulted in significantly different aerobic granule characteristics. Over 30 days, the chemical oxygen demand (COD) removal of the two reactors was maintained at similar levels (above 96%), and typical bioflocs were not observed. The average aerobic granule size in R2 was twice that in R1, as settling velocity increased in proportion to size increment. Meanwhile, the production yields of polysaccharide and protein content in R2 were always higher than those in R1. However, due to mass transfer limitations and the presence of anaerobes in the aerobic granule cores, larger granules had a tendency to disintegrate in R2. Thus, we conclude that a sludge discharge port situated at the reactor bottom is beneficial for aerobic granule stability, and enhances the potential for long-term aerobic granule SBR operation.

  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. 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.

  13. Utilizing a Russian Space Nuclear Reactor for a United States Space Mission: Systems Integration Issues

    NASA Astrophysics Data System (ADS)

    Reynolds, Edward; Schaefer, Edward; Polansky, Gary; Lacy, Jeff; Bocharov, Anatoly

    1994-07-01

    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 U.S. 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 learned 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. 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

  15. Multiplying steady-state culture in multi-reactor system.

    PubMed

    Erm, Sten; Adamberg, Kaarel; Vilu, Raivo

    2014-11-01

    Cultivation of microorganisms in batch experiments is fast and economical but the conditions therein change constantly, rendering quantitative data interpretation difficult. By using chemostat with controlled environmental conditions the physiological state of microorganisms is fixed; however, the unavoidable stabilization phase makes continuous methods resource consuming. Material can be spared by using micro scale devices, which however have limited analysis and process control capabilities. Described herein are a method and a system combining the high throughput of batch with the controlled environment of continuous cultivations. Microorganisms were prepared in one bioreactor followed by culture distribution into a network of bioreactors and continuation of independent steady state experiments therein. Accelerostat cultivation with statistical analysis of growth parameters demonstrated non-compromised physiological state following distribution, thus the method effectively multiplied steady state culture of microorganisms. The theoretical efficiency of the system was evaluated in inhibitory compound analysis using repeated chemostat to chemostat transfers.

  16. 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.

  17. 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.

  18. Summary of the fourth conference on United States utility experience in reactor noise analysis

    SciTech Connect

    Fry, D.N.

    1987-01-01

    The fourth informal conference on United States utility experience in reactor noise analysis and loose-part monitoring was held at the Northeast Utilities Service Company offices in Hartford, Connecticut, May 12-14, 1987. Host and general chairman for the meeting was J.V. Persio of Northeast Utilities. This conference provided a forum where utilities could share information on reactor noise analysis on an informal basis. There were about 60 attendees at the meeting representing 10 utilities, 3 reactor vendors, 8 consulting organizations, and 4 universities and research laboratories. Twenty-three papers were presented at the conference, dealing with various aspects of loose-part monitoring, neutron noise analysis, and standards activities.

  19. An ultracold neutron source at the NC State University PULSTAR reactor

    NASA Astrophysics Data System (ADS)

    Korobkina, E.; Wehring, B. W.; Hawari, A. I.; Young, A. R.; Huffman, P. R.; Golub, R.; Xu, Y.; Palmquist, G.

    2007-08-01

    Research and development is being completed for an ultracold neutron (UCN) source to be installed at the PULSTAR reactor on the campus of North Carolina State University (NCSU). The objective is to establish a university-based UCN facility with sufficient UCN intensity to allow world-class fundamental and applied research with UCN. To maximize the UCN yield, a solid ortho-D 2 converter will be implemented coupled to two moderators, D 2O at room temperature, to thermalize reactor neutrons, and solid CH 4, to moderate the thermal neutrons to cold-neutron energies. The source assembly will be located in a tank of D 2O in the space previously occupied by the thermal column of the PULSTAR reactor. Neutrons leaving a bare face of the reactor core enter the D 2O tank through a 45×45 cm cross-sectional area void between the reactor core and the D 2O tank. Liquid He will cool the disk-shaped UCN converter to below 5 K. Independently, He gas will cool the cup-shaped CH 4 cold-neutron moderator to an optimum temperature between 20 and 40 K. The UCN will be transported from the converter to experiments by a guide with an inside diameter of 16 cm. Research areas being considered for the PULSTAR UCN source include time-reversal violation in neutron beta decay, neutron lifetime determination, support measurements for a neutron electric-dipole-moment search, and nanoscience applications.

  20. 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.

  1. Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

    NASA Astrophysics Data System (ADS)

    Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

    2017-01-01

    In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

  2. 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

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION License Amendment Request From The State University of New York, University of Buffalo Reactor Facility AGENCY: Nuclear Regulatory Commission. ACTION: Notice of decommissioning plan, proposed...

  3. Can high fields save the tokamak? The challenge of steady-state operation for low cost compact reactors

    NASA Astrophysics Data System (ADS)

    Freidberg, Jeffrey; Dogra, Akshunna; Redman, William; Cerfon, Antoine

    2016-10-01

    The development of high field, high temperature superconductors is thought to be a game changer for the development of fusion power based on the tokamak concept. We test the validity of this assertion for pilot plant scale reactors (Q 10) for two different but related missions: pulsed operation and steady-state operation. Specifically, we derive a set of analytic criteria that determines the basic design parameters of a given fusion reactor mission. As expected there are far more constraints than degrees of freedom in any given design application. However, by defining the mission of the reactor under consideration, we have been able to determine the subset of constraints that drive the design, and calculate the values for the key parameters characterizing the tokamak. Our conclusions are as follows: 1) for pulsed reactors, high field leads to more compact designs and thus cheaper reactors - high B is the way to go; 2) steady-state reactors with H-mode like transport are large, even with high fields. The steady-state constraint is hard to satisfy in compact designs - high B helps but is not enough; 3) I-mode like transport, when combined with high fields, yields relatively compact steady-state reactors - why is there not more research on this favorable transport regime?

  4. Microwave-induced plasma reactor based on a domestic microwave oven for bulk solid state chemistry

    SciTech Connect

    Brooks, David J.; Douthwaite, Richard E.

    2004-12-01

    A microwave-induced plasma (MIP) reactor has been constructed from a domestic microwave oven (DMO) and applied to the bulk synthesis of solid state compounds. Low pressure MIP can be initiated and maintained using a range of gases including Ar, N{sub 2}, NH{sub 3}, O{sub 2}, Cl{sub 2}, and H{sub 2}S. In order to obtain reproducible synthesis conditions the apparatus is designed to allow control of gas flow rate, gas composition, and pressure. The use of the reactor is demonstrated by the synthesis of three binary metal nitrides formed in a NH{sub 3} MIP. The reactions are rapid and the products show good crystallinity and phase purity as judged by powder x-ray diffraction.

  5. 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.

  6. Neutron depth profiling at the University of Texas research reactor

    SciTech Connect

    Unlu, K.; Wehring, B.W. )

    1993-01-01

    A neutron depth profiling (NDP) facility has been developed at the University of Texas at Austin (UT) Nuclear Engineering Teaching Laboratory. The UT-NDP utilizes thermal neutrons from a tangential beam port of the 1-MW TRIGA Mark II research reactor. Aspects of the designs of the thermal neutron beam and target chamber for the UT-NDP facility are given in this paper. Also, a brief description of NDP and possible applications are included.

  7. 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

  8. 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.

  9. 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.

  10. Li-Ion Batteries for Forensic Neutron Dosimetry

    DTIC Science & Technology

    2016-03-01

    and a battery from the consumer electronic device (Cell03) were exposed. The batteries were exposed at two facilities, the TRIGA reactor at Oregon ...at TRIGA for comparison. 21 Figure 13: The spectra of the TRIGA ( Oregon State) and the FNI (UMass Lowell) [7] facilities. Note that the...simulation can be completed. The LIR2032 (CC01) and iPod shuffle battery (Cell03) were irradiated in the Oregon State University TRIGA reactor. Whereas

  11. Plasma flow reactor for steady state monitoring of physical and chemical processes at high temperatures

    NASA Astrophysics Data System (ADS)

    Koroglu, Batikan; Mehl, Marco; Armstrong, Michael R.; Crowhurst, Jonathan C.; Weisz, David G.; Zaug, Joseph M.; Dai, Zurong; Radousky, Harry B.; Chernov, Alex; Ramon, Erick; Stavrou, Elissaios; Knight, Kim; Fabris, Andrea L.; Cappelli, Mark A.; Rose, Timothy P.

    2017-09-01

    We present the development of a steady state plasma flow reactor to investigate gas phase physical and chemical processes that occur at high temperature (1000 < T < 5000 K) and atmospheric pressure. The reactor consists of a glass tube that is attached to an inductively coupled argon plasma generator via an adaptor (ring flow injector). We have modeled the system using computational fluid dynamics simulations that are bounded by measured temperatures. In situ line-of-sight optical emission and absorption spectroscopy have been used to determine the structures and concentrations of molecules formed during rapid cooling of reactants after they pass through the plasma. Emission spectroscopy also enables us to determine the temperatures at which these dynamic processes occur. A sample collection probe inserted from the open end of the reactor is used to collect condensed materials and analyze them ex situ using electron microscopy. The preliminary results of two separate investigations involving the condensation of metal oxides and chemical kinetics of high-temperature gas reactions are discussed.

  12. 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…

  13. 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.

  14. 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.

  15. Computer simulation of reactions in beta-cyclodextrin molecular reactors: transition state recognition.

    PubMed

    Yeguas, Violeta; López, Ramón; Lambert, Alexandrine; Monard, Gérald; Ruiz-López, Manuel F

    2010-10-07

    Cyclodextrins have attracted much interest in recent years because of their potential use as molecular reactors allowing organic reactions in aqueous solution. To better understand their effect on reaction mechanisms, we have carried out a computational study of a prototypical process (neutral ester hydrolysis) in a beta-cyclodextrin (beta-CD). Two models have been used for the reactor. The first and simpler one assumes that the medium can be described by a polarizable dielectric continuum. The second one takes into account the discrete nature of the beta-CD and water molecules thanks to a computational approach that combines the use of Quantum Mechanics, Molecular Mechanics and Molecular Dynamics techniques. We focus on neutral pH processes for which either acceleration or inhibition has experimentally been observed depending on ester derivatives. Our calculations rationalize such observations by showing that the two reaction mechanisms usually invoked for hydrolysis, stepwise (involving two transitions states with formation of a -C(OH)(2)OR tetrahedral intermediate) and concerted, undergo opposite effects in the beta-CD environment. The results highlight the role played by molecular shape recognition. Thus, in spite of a higher polarity exhibited by the three transition states with respect to the reactants, the interactions with the beta-CD cavity may either increase or decrease the activation barrier due to different 3D-arrangements of the chemical structures.

  16. A simplified model for the steady-state biofilm-activated sludge reactor.

    PubMed

    Fouad, Moharram; Bhargava, Renu

    2005-02-01

    A simplified mathematical model is proposed to describe the steady-state completely mixed biofilm-activated sludge reactor (hybrid reactor). The model is derived based on Monod kinetic expressions and the Fickian diffusion law in biofilm. In addition, it considers all the essential concepts that describe the two types of growth (suspended and attached) and the competition between them for limiting substrate. Also the present study has been extended to investigate simple and accurate mathematical expressions for describing the substrate diffusion in biofilm (J). The expression for substrate flux has an explicit solution, which may be useful in the proposed model and many other applications. The application of the model for the hybrid system has been explained for a given set of data and verified by comparison with another solution. Also the model was applied to experimental results for a trace level of suspended biomass concentration (X). It was found that the biofilm flux (J) is the key factor in the model prediction, hence the accuracy of the model output is influenced by the accuracy of J. Compared with other solutions for such systems the model is simple, easy to use, and provides an accurate tool for describing such systems based on fundamental principles.

  17. Development of an Ultracold Neutron Source at the NC State PULSTAR Reactor

    NASA Astrophysics Data System (ADS)

    Palmquist, Grant R.; Golub, Robert; Hawari, Ayman I.; Holley, Adam T.; Huffman, Paul R.; Korobkina, Ekaterina; Wehring, Bernard W.; Xu, Yanping; Young, Albert R.

    2006-10-01

    Development of an ultracold neutron (UCN) source is underway at the North Carolina State University 1 MW PULSTAR reactor facility. Fast neutrons from the reactor core are moderated first by heavy water at room temperature and then by solid methane at a temperature of approximately 25 K. The cold neutrons emerging from the methane are then down-scattered to lower energies (< 300 neV) by phonons in the 5 K solid ortho-deuterium converter. The UCN emerging from the deuterium source are guided to the experimental area through diamond-like carbon-coated quartz. The anticipated UCN densities in an experimental volume connected to the end of the guide should exceed densities presently obtained at the UCN source at the Institute Laue--Langevin. The current status of the design and construction will be discussed, with an emphasis on the thermal modeling of the cryostat cooling system. This work is supported in part by NSF grant #0314114 and funds from the DOE INIE program.

  18. 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.

  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. Chemical Reactors.

    ERIC Educational Resources Information Center

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  1. Advanced fueling system for steady-state operation of a fusion reactor

    SciTech Connect

    Raman, R.

    2008-07-15

    Steady-state Advanced Tokamak scenarios rely on optimized density and pressure profiles to maximize the bootstrap current fraction. Under this mode of operation, the fuelling system must deposit small amounts of fuel where it is needed, and as often as needed, so as to compensate for fuel losses, but not to adversely alter the established density and pressure profiles. A precision fuelling system has the capability for controlling the fusion burn by maintaining the required pressure profile to maximize the bootstrap current fraction. An advanced fuelling system based on Compact Toroid (CT) injection has the potential to meet these needs while simultaneously simplifying the requirements of the tritium handling systems. Simpler engineering systems would reduce reactor construction and maintenance cost through increased reliability. A CT fueling system is described together with the associated tritium handling requirements. (authors)

  2. 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.

  3. 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).

  4. NEUTRON AND GAMMA DOSIMETRY MEASUREMENTS AT THE AFRRI=DASA TRIGA REACTOR.

    DTIC Science & Technology

    IONIZATION CHAMBERS, THERMAL NEUTRONS, FOILS(MATERIALS), RADIATION MEASURING INSTRUMENTS, PLUTONIUM, NEPTUNIUM , URANIUM, SULFUR, GRAPHITE, SILVER COMPOUNDS, PHOSPHATES, CARBON DIOXIDE, CADMIUM, GOLD, INDIUM.

  5. Fission products measured from highly-enriched uranium irradiated under 10B4C in a research reactor

    SciTech Connect

    Metz, Lori A.; Friese, Judah I.; Finn, Erin C.; Greenwood, Lawrence R.; Hines, Corey C.; King, Matthew D.; Wall, Donald E.

    2016-03-01

    Critical assemblies provide one method of achieving a fast neutron spectrum that is close to a 235U fission-energy neutron spectrum for nuclear data measurements. Previous work has demonstrated the use of a natural boron carbide capsule for spectral-tailoring in a mixed spectrum reactor as an alternate and complementary method for performing fission-energy neutron experiments. Previous fission products measurements showed that the neutron spectrum achievable with natural boron carbide was not as hard as what can be achieved with critical assemblies. New measurements performed with the Washington State University TRIGA reactor using a boron carbide capsule 96% enriched in 10B for irradiations resulted in a neutron spectrum very similar to a critical assembly and a pure 235U fission spectrum. The current work describes an experiment involving a highly-enriched uranium target irradiated under the new 10B4C capsule. Fission product yields were measured following radiochemical separations and are presented here. Reactor dosimetry measurements for characterizing neutron spectra and fluence for the enriched boron carbide capsule and critical assemblies are also discussed.

  6. Plutonium distribution and oxidation states in a reactor leaching ponds system

    SciTech Connect

    Ibrahim, S.; Culp, T. )

    1989-10-01

    Concentrations of 239,240Pu and 238Pu in water, net plankton (algal material), suspended particulates and sediment, as well as Pu oxidation states in filtered water, were determined in a test reactor leaching ponds system in southeastern Idaho. The highest Pu concentration in the ponds system was found in net plankton, and concentrations varied significantly between sampling dates. Plutonium Concentration Ratios (CR) for plankton ranged from 3 X 10(4) to 4 X 10(5). The lowest Pu concentration was found in filtered water, primarily because of the absence of complexing agents. The majority of Pu in filtered water was in true solution (60-87%) or present in colloidal particles smaller than 0.22 micron. Plutonium association with sediment was inversely related to particle size. The environmental distribution coefficients (Kd) for Pu ranged from 1.6 X 10(4) to 1.2 X 10(5) reflecting the importance of sediments as the main reservoir for Pu in the ponds system. No significant differences were noted between CR or Kd values for 239,240Pu and 238Pu. The reduced Pu oxidation states (III and IV) fractions ranged from 57% to 71% of the total dissolved Pu in water. This is in contrast with oxidation states distribution from other large aquatic systems (Great Lakes and the Irish Sea) where Pu is predominately in oxidized (V and VI) forms.

  7. 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.

  8. 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

  9. 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.

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

    PubMed

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

    2015-10-09

    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.

  11. Survey of light-water-reactor designs to be offered in the United States

    SciTech Connect

    Spiewak, I.

    1986-03-01

    ORNL has conducted a Nuclear Power Options Viability Study for the Department of Energy. That study is primarily concerned with new technology which could be developed for initial operation in the 2000 to 2010 time frame. Such technology would have to compete not only with coal options but with incrementally improved commercial light-water-reactors. This survey reported here was undertaken to gain an understanding of the nuclear commercial technology likely to be offered in the late 1980s and perhaps beyond. The three US vendors actively marketing NSSSs are each developing a product for the future which they expect to be more reliable, more maintainable, more economical, and safer than the present plants. These are all essentially 3800-MW(t) designs, although all are studying smaller plants. They apparently will be off offered as standard prelicensed designs with much larger scope than earlier NSSS offerings, with the possibility of firm prices. Westinghouse with Mitsubishi Heavy Industries is developing a completely new design (APWR) to be built initially in Japan, hopefully for operation by the mid-1990s. Westinghouse is making a strong effort to have the APWR licensed in the US as a standard plant. Combustion Engineering (C-E) is evaluating potential improvements to the System-80 standard design (CESSAR) that has already received final design approval by the NRC. General Electric (GE), with Hitachi and Toshiba, is developing a new design (ABWR) that incorporates advanced features which have been proven by the worldwide BWR suppliers. The ABWR is to be built initially in Japan, but the design could be adapted to the United States. Westinghouse, C-E, and GE have done some conceptual evaluation of reactors in the 600-MW(e) class. The Westinghouse concept is a two-loop plant intended for factory assembly in a shipyard and delivery to a site by barge. The GE concept is a modification of the ABWR with some additional passive safety features. 16 figs.

  12. STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

    SciTech Connect

    Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart

    2016-05-01

    With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

  13. Use of tri-gas incubator for routine culture of Campylobacter species from fecal specimens.

    PubMed Central

    Thompson, J S; Hodge, D S; Smith, D E; Yong, Y A

    1990-01-01

    We evaluated a tri-gas incubator for Campylobacter isolation to be used instead of an anaerobic jar. Fecal specimens were cultured in duplicate onto charcoal selective medium and incubated at 43 degrees C for 48 h in two different environments: a tri-gas incubator (Forma Scientific) adjusted to provide an atmosphere of 10% CO2, 10% O2, and the balance N2; and evacuated anaerobic jars with a replacement gas mixture of 10% CO2, 5% O2, and 85% N2. A total of 106 Campylobacter jejuni and 8 Campylobacter coli isolates were obtained from 2,348 stool specimens. Of the positive specimens, 113 isolates came from the incubator and 111 isolates came from the anaerobic jars. An additional 32 previously positive specimens were replated onto charcoal selective medium and retested by both methods. We recovered 27 C. jejuni isolates, 26 isolates by each method. The isolates from the incubator typically produced discrete colonies, while the isolates from the anaerobic jar showed some degree of swarming in colony formation. The tri-gas incubator provided a cost-effective method for culturing Campylobacter spp. PMID:2280012

  14. 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.

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

    SciTech Connect

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

    2016-01-22

    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.

  16. 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.

  17. 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.

  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. )

    1992-01-01

    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.

  19. 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.

  20. State observers for a biological wastewater nitrogen removal process in a sequential batch reactor.

    PubMed

    Boaventura, K M; Roqueiro, N; Coelho, M A; Araújo, O Q

    2001-08-01

    Biological removal of nitrogen is a two-step process: aerobic autotrophic microorganisms oxidize ammoniacal nitrogen to nitrate, and the nitrate is further reduced to elementary nitrogen by heterotrophic microorganisms under anoxic condition with concomitant organic carbon removal. Several state variables are involved which render process monitoring a demanding task, as in most biotechnological processes, measurement of primary variables such as microorganism, carbon and nitrogen concentrations is either difficult or expensive. An alternative is to use a process model of reduced order for on-line inference of state variables based on secondary process measurements, e.g. pH and redox potential. In this work, two modeling approaches were investigated: a generic reduced order model based on the generally accepted IAWQ No. 1 Model [M. Henze, C.P.L., Grady, W., Gujer, G.V.R., Marais, T., Matsuo, Water Res. 21 (5) (1987) 505-515]-generic model (GM), and a reduced order model specially validated with the data acquired from a benchscale sequential batch reactor (SBR) specific model (SM). Model inaccuracies and measurement errors were compensated for with a Kalman filter structure to develop two state observers: one built with GM, the generic observer (GO), and another based on SM, the specific observer (SO). State variables estimated by GM, SM, GO and SO were compared to experimental data from the SBR unit. GM gave the worst performance while SM predictions presented some model to data mismatch. GO and SO, on the other hand, were both in very good agreement with experimental data showing that filters add robustness against model errors, which reduces the modeling effort while assuring adequate inference of process variables.

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

    NASA Astrophysics Data System (ADS)

    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 alinement 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.

  2. 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

    2016-09-01

    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. 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. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).

  3. 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)

  4. 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

  5. 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

  6. Economic Impacts on the United States of Siting Decisions for the International Thermonuclear Experimental Reactor

    NASA Astrophysics Data System (ADS)

    Peerenboom, J. P.; Hanson, M. E.; Huddleston, J. R.; Wolsko, T. D.

    1997-12-01

    This paper 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.

  7. 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.

  8. Divertor erosion study for TPX and implications for steady-state fusion reactors

    SciTech Connect

    Brooks, J.N.

    1995-12-31

    A sputtering erosion analysis was performed for the tilted plate divertor design of the proposed TPX tokamak. High temperature ({approximately}100 eV), non-radiative, steady-state compatible, plasma edge conditions were used as input to the REDEP erosion/redeposition code. For the reference carbon surface the results show a stable erosion profile, i.e., non-runaway self-sputtering, in spite of carbon self-sputtering coefficients that are locally in excess of unity. The resulting net erosion rates are high (peak {approx}1--2.5 m/burn-yr) but may be acceptable for a low duty factor experimental device such as TPX. Other surface materials were also analyzed, in part to obtain insight for fusion reactor designs using a similar plasma regime. Both medium and high-Z materials are predicted not to work, due to runaway self-sputtering. Beryllium is stable but has erosion rates as high or higher than carbon. A liquid metal lithium surface has stable sputtering with a zero-erosion potential and may thus be an attractive future material choice.

  9. Divertor erosion study for TPX and implications for steady-state fusion reactors

    SciTech Connect

    Brooks, J.N.

    1995-12-31

    A sputtering erosion analysis was performed for the tilted plate divertor design of the proposed TPX tokamak. High temperature ({approximately} 100 eV), non-radiative, steady-state compatible, plasma edge conditions were used as input to the REDEP erosion/redeposition code. For the reference carbon surface the results show a stable erosion profile, i.e., non-runaway self-sputtering, in spite of carbon self-sputtering coefficients that are locally in excess of unity. The resulting net erosion rates are high (peak {approx} 1--2.5 m/burn-yr) but may be acceptable for a low duty factor experimental device such as TPX. Other surface materials were also analyzed, in part to obtain insight for fusion reactor designs using a similar plasma regime. Both medium and high-Z materials are predicted not to work, due to runaway self-sputtering. Beryllium is stable but has erosion rates as high or higher than carbon. A liquid metal lithium surface has stable sputtering with a zero-erosion potential and may thus be an attractive future material choice.

  10. 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)

  11. 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.

  12. 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.

  13. 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.

  14. 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.)

  15. Dynamic simulation platform to verify the performance of the reactor regulating system for a research reactor

    SciTech Connect

    2015-07-01

    Digital instrumentation and controls system technique is being introduced in new constructed research reactor or life extension of older research reactor. Digital systems are easy to change and optimize but the validated process for them is required. Also, to reduce project risk or cost, we have to make it sure that configuration and control functions is right before the commissioning phase on research reactor. For this purpose, simulators have been widely used in developing control systems in automotive and aerospace industries. In these literatures, however, very few of these can be found regarding test on the control system of research reactor with simulator. Therefore, this paper proposes a simulation platform to verify the performance of RRS (Reactor Regulating System) for research reactor. This simulation platform consists of the reactor simulation model and the interface module. This simulation platform is applied to I and C upgrade project of TRIGA reactor, and many problems of RRS configuration were found and solved. And it proved that the dynamic performance testing based on simulator enables significant time saving and improves economics and quality for RRS in the system test phase. (authors)

  16. Compact Reactor

    NASA Astrophysics Data System (ADS)

    Williams, Pharis E.

    2007-01-01

    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.

  17. 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.

  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. Design of a reactor inlet temperature controller for EBR-2 using state feedback

    SciTech Connect

    Vilim, R.B.; Planchon, H.P.

    1990-01-01

    A new reactor inlet temperature controller for pool type liquid-metal reactors has been developed and will be tested in EBR-II. The controller makes use of modern control techniques to take into account stratification and mixing in the cold pool during normal operation. Secondary flowrate is varied so that the reactor inlet temperature tracks a setpoint while reactor outlet temperature, primary flowrate and secondary cold leg temperature are treated as exogenous disturbances and are free to vary. A disturbance rejection technique minimizes the effect of these disturbances on inlet temperature. A linear quadratic regulator improves inlet temperature response. Tests in EBR-II will provide experimental data for assessing the performance improvements that modern control can produce over the existing EBR-II analog inlet temperature controller. 10 refs., 8 figs.

  20. Analyses of the transportation of spent research reactor fuel in the United States

    SciTech Connect

    Cashwell, J.W.; Neuhauser, K.S.

    1989-01-01

    We analyzed the impacts of transportation of research reactor spent fuel from US and foreign reactors for the US Department of Energy's (DOE) Office of Defense Programs. Two separate shipment programs were analyzed. The shipment of research reactor spent fuel from Taiwan to the US (Fuel Movement Program), and the return of research reactor spent fuels of US origin from foreign and domestic reactors (Research Reactor Fuel Return Program). To perform these analyses, a comprehensive methodology for analyzing the probabilities and consequences of transportation in coastal waters and port facilities, handling at the port, and shipment by truck to reprocessing facilities was developed. The Taiwanese fuel consists of low-burnup aluminum-clad metallic uranium research reactor spent fuel; the other fuels are primarily aluminum-clad oxide fuels. The Fuel Movement Program is ongoing, while the Fuel Return Program addresses future shipments over a ten-year period. The operational aspects of the Taiwanese shipments have been uniform, but several possible shipping configurations are possible for the Fuel Return Program shipments. The risks of transporting spent nuclear fuel and other radioactive materials by all modes have been analyzed extensively. Comprehensive assessments, which bound the impacts of spent fuel transport, demonstrate that when shipments are made in compliance with applicable regulations, the risks for all such transport are low. For comparison with previously licensed transport activities and to provide continuity with earlier analyses, the results for shipment of 150-day-old commercial pressurized water reactor (PWR) spent fuel are presented as part of this study.

  1. Impact of recent developments on the status of commercial nuclear power and reactor decommissioning in the United States

    SciTech Connect

    Devgun, Jas S.

    2013-07-01

    This paper examines the impact of three recent developments on the commercial nuclear power in the United States. These developments include: Yucca Mountain closure and issues related to SNF; actions in response to Fukushima Diaiichi accident, and; energy economics. All of these have had a significant impact on the commercial nuclear power, its future, as well as the reactor decommissioning scene in the US. (authors)

  2. Utilization of the Cornell University research reactors in support of the Nuclear Power Industry

    SciTech Connect

    Aderhold, A.C. )

    1993-01-01

    Cornell University is licensed to operate two research reactor facilities on its main campus in Ithaca, New York: a 500-kW pulsing TRIGA and a 100-W zero-power reactor (ZPR). The initial criticality of both reactors took place in 1962, and the utilization of each has been, and continues to be, dedicated to the teaching and research programs of Cornell's many academic departments. As the nation's nuclear power industry grew, the demand for services at research and test reactors increased. As a result, and in large part because of special design features of the TRIGA, Cornell responded to a few requests for reactor testing services while maintaining the policy that these services would not interfere with teaching and research programs. The frequency of service requests suddenly mushroomed in November of 1989, when the nation's major testing reactor was shut down for repairs. In spite of a small staff of two full-time reactor operators, a decision was made to support the nuclear industry to the fullest extent possible without jeopardizing Cornell's teaching and research programs. This turned into a monumental task of tight scheduling and meeting precise deadlines. It could only be accomplished by working late evenings and weekends and, on a number of occasions, staying at the facility for up to 5 days continuously.

  3. The preliminary analysis on the steady-state and kinetic features of the molten salt pebble-bed reactor

    SciTech Connect

    Xia, B.; Lu, Y.

    2012-07-01

    A novel design concept of molten salt pebble-bed reactor with an ultra-simplified integral primary circuit called 'Nuclear Hot Spring' has been proposed, featured by horizontal coolant flow in a deep pool pebble-bed reactor, providing 'natural safety' features with natural circulation under full power operation and less expensive primary circuit arrangement. In this work, the steady-state physical properties of the equilibrium state of the molten salt pebble-bed reactor are calculated by using the VSOP code, and the steady-state thermo-hydraulic analysis is carried out based on the approximation of absolutely horizontal flow of the coolant through the core. A new concept of 2-dimensional, both axial and radial, multi-pass on-line fuelling scheme is presented. The result reveals that the radial multi-pass scheme provides more flattened power distribution and safer temperature distribution than the one-pass scheme. A parametric analysis is made corresponding to different pebble diameters, the key parameter of the core resistance and the temperature at the pebble center. It is verified that within a wide range of pebble diameters, the maximum pebble center temperatures are far below the safety limit of the fuel, and the core resistance is considerably less than the buoyant force, indicating that the natural circulation under full power operation is achievable and the ultra-simplified integral primary circuit without any pump is possible. For the kinetic properties, it is verified that the negative temperature coefficient is achieved in sufficient under-moderated condition through the preliminary analysis on the temperature coefficients of fuel, coolant and moderator. The requirement of reactivity compensation at the shutdown stages of the operation period is calculated for the further studies on the reactivity control. The molten salt pebble-bed reactor with horizontal coolant flow can provide enhanced safety and economical features. (authors)

  4. Review of the state of criticality of the Three Mile Island Unit 2 core and reactor vessel

    SciTech Connect

    Stratton, W.R. )

    1987-04-15

    The events during the early hours of the Three Mile Island Unit 2 (TMI-2) accident on March 28, 1979 caused the fuel in the reactor core to crumble or disintegrate, and then subside into a rubble structure more compact that its normal configuration. The present height of the core is about seven feet, five feet less than its normal configuration of 12 feet. With the same boron content and some or all of the control rod and burnable poison rod material as the normal core configuration, the collapsed structure is calculated to be more reactive. However, the reactor is assuredly subcritical at present because of the extraordinarily high boron concentration maintained in the coolant water. Four additional and different physical models are discussed briefly in the report to illustrate the margin of subcriticality, to provide a better estimate of the neutron multiplication factor, and to provide some understanding of the criticality effects of the important parameters. Two different finite, cylindrical models of a collapsed core are also presented in this report. The conclusion of this review is that the reactor is now very far subcritical with a boron concentration of 4350 ppM or more, and no conceivable rearrangement of fuel can create a critical state. Careful administrative control to maintain the boron concentration of the reactor coolant close to 5000 ppM, and controls to rigorously exclude addition of unborated water to the primary system, provide additional assurance that subcriticality will be maintained. The immediate corollary is that the defueling of the reactor vessel can proceed as planned, with complete confidence that such operations will remain subcritical. 20 refs.

  5. 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.

  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. Tokamak reactor studies

    SciTech Connect

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features.

  8. 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.

  9. High internal inductance for steady-state operation in ITER and a reactor

    SciTech Connect

    Ferron, John R.; Holcomb, Christopher T.; Luce, Timothy C.; Park, J. M.; Kolemen, Egemen; La Haye, Robert J.; Solomon, Wayne M.; Turco, Francesca

    2015-06-26

    Increased confinement and ideal stability limits at relatively high values of the internal inductance (${{\\ell}_{i}}$ ) have enabled an attractive scenario for steady-state tokamak operation to be demonstrated in DIII-D. Normalized plasma pressure in the range appropriate for a reactor has been achieved in high elongation and triangularity double-null divertor discharges with ${{\\beta}_{\\text{N}}}\\approx 5$ at ${{\\ell}_{i}}\\approx 1.3$ , near the ideal $n=1$ kink stability limit calculated without the effect of a stabilizing vacuum vessel wall, with the ideal-wall limit still higher at ${{\\beta}_{\\text{N}}}>5.5$ . Confinement is above the H-mode level with ${{H}_{98\\left(\\text{y},2\\right)}}\\approx 1.8$ . At ${{q}_{95}}\\approx 7.5$ , the current is overdriven, with bootstrap current fraction ${{f}_{\\text{BS}}}\\approx 0.8$ , noninductive current fraction ${{f}_{\\text{NI}}}>1$ and negative surface voltage. For ITER (which has a single-null divertor shape), operation at ${{\\ell}_{i}}\\approx 1$ is a promising option with ${{f}_{\\text{BS}}}\\approx 0.5$ and the remaining current driven externally near the axis where the electron cyclotron current drive efficiency is high. This scenario has been tested in the ITER shape in DIII-D at ${{q}_{95}}=4.8$ , so far reaching ${{f}_{\\text{NI}}}=0.7$ and ${{f}_{\\text{BS}}}=0.4$ at ${{\\beta}_{\\text{N}}}\\approx 3.5$ with performance appropriate for the ITER Q=5 mission, ${{H}_{89}}{{\\beta}_{\\text{N}}}/q_{95}^{2}\\approx 0.3$ . Modeling studies explored how increased current drive power for DIII-D could be applied to maintain a stationary, fully noninductive high ${{\\ell}_{i}}$ discharge. Lastly, stable solutions in the double-null shape are found without the vacuum vessel wall at ${{\\beta}_{\\text{N}}}=4$ , ${{\\ell}_{i}}=1.07$ and ${{f}_{\\text{BS}}}=0.5$ , and at ${{\\beta}_{\\text{N}}}=5$ with the vacuum vessel wall.

  10. High internal inductance for steady-state operation in ITER and a reactor

    DOE PAGES

    Ferron, John R.; Holcomb, Christopher T.; Luce, Timothy C.; ...

    2015-06-26

    Increased confinement and ideal stability limits at relatively high values of the internal inductance (more » $${{\\ell}_{i}}$$ ) have enabled an attractive scenario for steady-state tokamak operation to be demonstrated in DIII-D. Normalized plasma pressure in the range appropriate for a reactor has been achieved in high elongation and triangularity double-null divertor discharges with $${{\\beta}_{\\text{N}}}\\approx 5$$ at $${{\\ell}_{i}}\\approx 1.3$$ , near the ideal $n=1$ kink stability limit calculated without the effect of a stabilizing vacuum vessel wall, with the ideal-wall limit still higher at $${{\\beta}_{\\text{N}}}>5.5$$ . Confinement is above the H-mode level with $${{H}_{98\\left(\\text{y},2\\right)}}\\approx 1.8$$ . At $${{q}_{95}}\\approx 7.5$$ , the current is overdriven, with bootstrap current fraction $${{f}_{\\text{BS}}}\\approx 0.8$$ , noninductive current fraction $${{f}_{\\text{NI}}}>1$$ and negative surface voltage. For ITER (which has a single-null divertor shape), operation at $${{\\ell}_{i}}\\approx 1$$ is a promising option with $${{f}_{\\text{BS}}}\\approx 0.5$$ and the remaining current driven externally near the axis where the electron cyclotron current drive efficiency is high. This scenario has been tested in the ITER shape in DIII-D at $${{q}_{95}}=4.8$$ , so far reaching $${{f}_{\\text{NI}}}=0.7$$ and $${{f}_{\\text{BS}}}=0.4$$ at $${{\\beta}_{\\text{N}}}\\approx 3.5$$ with performance appropriate for the ITER Q=5 mission, $${{H}_{89}}{{\\beta}_{\\text{N}}}/q_{95}^{2}\\approx 0.3$$ . Modeling studies explored how increased current drive power for DIII-D could be applied to maintain a stationary, fully noninductive high $${{\\ell}_{i}}$$ discharge. Lastly, stable solutions in the double-null shape are found without the vacuum vessel wall at $${{\\beta}_{\\text{N}}}=4$$ , $${{\\ell}_{i}}=1.07$$ and $${{f}_{\\text{BS}}}=0.5$$ , and at $${{\\beta}_{\\text{N}}}=5$$ with the vacuum vessel wall.« less

  11. Fuel burnup analysis for Thai research reactor by using MCNPX computer code

    NASA Astrophysics Data System (ADS)

    Sangkaew, S.; Angwongtrakool, T.; Srimok, B.

    2017-06-01

    This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.

  12. 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.

  13. Microbial population dynamics of granular aerobic sequencing batch reactors during start-up and steady state periods.

    PubMed

    Liu, Y Q; Kong, Y H; Zhang, R; Zhang, X; Wong, F S; Tay, J H; Zhu, J R; Jiang, W J; Liu, W T

    2010-01-01

    This study investigates microbial population dynamics in granular sequencing batch reactors (GSBR). The experimental results of DGGE fingerprint of sludge demonstrated that the microbial community structure of sludge shifted significantly during granulation period and nutrient removal improvement period. After reactor performance and physical characteristics of sludge reached steady state, microbial population of sludge became relatively stable. The high similarity of microbial community structure between co-existed flocculated sludge and granular sludge in GSBR at different operation phases indicated that similar microbial consortium could exist in compact aggregated form or in amorphous flocculated form. Therefore, strong selection pressure was still required to wash out flocs to maintain the stability of reactor operation. In addition, it was found that substrate type had considerable impact on microbial species selection and enrichment in granular sludge. The clone library of granular sludge showed that microbial species in divisions of α-Proteobacteria, β-Proteobacteria, γ-Proteobacteria and Bacteroidetes existed within acetate-fed granule communities and Thauera spp. from β-Proteobacteria accounted for 49% of the total clones in the whole clone library. It is thus speculated that Thauera spp. are important for the formation of acetate-fed granules under the conditions used in this study, maintaining the integrity of granules or substrate degradation.

  14. Physics-Based Multi-State Models of Passive Component Degradation for the R7 Reactor Simulation Environment

    SciTech Connect

    Unwin, Stephen D.; Layton, Robert F.; Johnson, Kenneth I.; Lowry, Peter P.

    2012-06-25

    Abstract: The Next Generation Systems Analysis Code - referred to as R7 - is reactor systems simulation software being developed to support the Risk-Informed Safety Margin Characterization Pathway of the U.S. Department of Energy's Light Water Reactor Sustainability Program. It will provide an integrated multi-physics environment, implemented in an uncertainty quantification (UQ) framework that can produce risk and other performance insights on long-term reactor operations. An element of this simulation environment will be the performance of passive components and materials. Conventional models of component reliability are largely parametric, relying on plant service data to estimate component lifetimes and failure rates. This type of model has limited usefulness in the R7 environment where the intent is to explicitly determine the influence of physical stressors on component degradation. In this paper, we describe a new class of multi-state physics-based component models designed to be R7-compatible. These models capture the physics of materials degradation while also incorporating the effects of interventions and component rejuvenation. The models are implemented in a cumulative damage framework that allows the impact of an evolving physical environment to be addressed without recourse to resampling within the Monte Carlo-based UQ framework. The paper describes an application to stress corrosion cracking in dissimilar metal welds - a principal contributor to potential loss of coolant accidents. So while R7 will have the more conventional capability of reactor simulation codes to model the impact of degraded components and systems on plant performance, the methodology described here allows R7 to model the inverse effect; the impact of the physical environment on component degradation and performance.

  15. Background studies for the MINER Coherent Neutrino Scattering reactor experiment

    NASA Astrophysics Data System (ADS)

    Agnolet, G.; Baker, W.; Barker, D.; Beck, R.; Carroll, T. J.; Cesar, J.; Cushman, P.; Dent, J. B.; De Rijck, S.; Dutta, B.; Flanagan, W.; Fritts, M.; Gao, Y.; Harris, H. R.; Hays, C. C.; Iyer, V.; Jastram, A.; Kadribasic, F.; Kennedy, A.; Kubik, A.; Lang, K.; Mahapatra, R.; Mandic, V.; Marianno, C.; Martin, R. D.; Mast, N.; McDeavitt, S.; Mirabolfathi, N.; Mohanty, B.; Nakajima, K.; Newhouse, J.; Newstead, J. L.; Ogawa, I.; Phan, D.; Proga, M.; Rajput, A.; Roberts, A.; Rogachev, G.; Salazar, R.; Sander, J.; Senapati, K.; Shimada, M.; Soubasis, B.; Strigari, L.; Tamagawa, Y.; Teizer, W.; Vermaak, J. I. C.; Villano, A. N.; Walker, J.; Webb, B.; Wetzel, Z.; Yadavalli, S. A.

    2017-05-01

    The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5-20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

  16. 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.

  17. 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.

  18. Calculation to experiment comparison of SPND signals in various nuclear reactor environments

    SciTech Connect

    Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien; Snoj, Luka; Tarchalski, Mikolaj; Dewynter-Marty, Veronique; Malouch, Fadhel

    2015-07-01

    In the perspective of irradiation experiments in the future Jules Horowitz Reactor (JHR), the Instrumentation Sensors and Dosimetry Laboratory of CEA Cadarache (France) is developing a numerical tool for SPND design, simulation and operation. In the frame of the SPND numerical tool qualification, dedicated experiments have been performed both in the Slovenian TRIGA Mark II reactor (JSI) and very recently in the French CEA Saclay OSIRIS reactor, as well as a test of two detectors in the core of the Polish MARIA reactor (NCBJ). A full description of experimental set-ups and neutron-gamma calculations schemes are provided in the first part of the paper. Calculation to experiment comparison of the various SPNDs in the different reactors is thoroughly described and discussed in the second part. Presented comparisons show promising final results. (authors)

  19. Diamond as a solid state micro-fission chamber for thermal neutron detection at the VR-1 research reactor

    SciTech Connect

    Pomorski, Michal; Mer-Calfati, Christine; Foulon, Francois; Sklenka, Lubomir; Rataj, Jan; Bily, Tomas

    2015-07-01

    Diamond exhibits a combination of properties which makes it attractive for neutron detection in hostile conditions. In the particular case of detection in a nuclear reactor, it is resilient to radiation, exhibits a natural low sensitivity to gamma rays, and its small size (as compared with that of gas ionisation chambers) enables fluency monitoring with a high position resolution. We report here on the use of synthetic CVD diamond as a solid state micro-fission chamber with U-235 converting material for in-core thermal neutron monitoring. Two types of thin diamond detectors were developed for this application. The first type of detector is fabricated using thin diamond membrane obtained by etching low-cost commercially available single crystal CVD intrinsic diamond, so called 'optical grade' material. Starting from a few hundred of micrometre thick samples, the sample is sliced with a laser and then plasma etched down to a few tenths of micrometre. Here we report the result obtained with a 17 μm thick device. The detection surface of this detector is equal to 1 mm{sup 2}. Detectors with surfaces up to 1 cm{sup 2} can be fabricated with this technique. The second type of detector is fabricated by growing successively two thin films of diamond, by the microwave enhanced chemical vapour deposition technique, on HPHT single crystal diamond. A first, a film of boron doped (p+) single crystal diamond, a few microns thick, is deposited. Then a second film of intrinsic diamond with a thickness of a few tens of microns is deposited. This results in a P doped, Intrinsic, Metal structure (PIM) structure in which the intrinsic volume id the active part of the detector. Here we report the results obtained with a 20 μm thick intrinsic whose detection surface is equal to 0.5 mm{sup 2}, with the possibility to enlarge the surface of the detector up to 1 cm{sup 2}. These two types of detector were tested at the VR-1 research reactor at the Czech Technical University in Prague

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Distribution of properties in nuclear reactor vessel shells in the unirradiated state

    NASA Astrophysics Data System (ADS)

    Skundin, M. A.; Chernobaeva, A. A.; Zhurko, D. A.; Krasikov, E. A.; Medvedev, K. I.

    2013-04-01

    The distributions of the chemical composition, the strength characteristics, and critical ductile-brittle transition temperature T cr are studied in the axial, radial, and tangential directions of the material of a test ring cut from a standard forging used for a VVER-1000 reactor vessel shell. The values of T cr of specimens cut from the test ring are shown to be well below those of the internal volume of the shell, which can explain the substantial scatter of the results obtained on reference specimens cut from the base metal.

  5. A carbon-cluster laser ion source for TRIGA-TRAP

    NASA Astrophysics Data System (ADS)

    Smorra, C.; Blaum, K.; Eberhardt, K.; Eibach, M.; Ketelaer, J.; Ketter, J.; Knuth, K.; Nagy, Sz

    2009-08-01

    A new laser ablation ion source was developed and tested for the Penning trap mass spectrometer TRIGA-TRAP in order to provide carbon-cluster ions for absolute mass calibration. Ions of different cluster sizes up to C+24 were successfully produced, covering the mass range up to the heavy actinide elements. The ions were captured in a Penning trap, and their time-of-flight cyclotron resonances recorded in order to determine their cyclotron frequency. Furthermore, the same ion source was used to produce GdO+ ions from a gadolinium target in sufficient amount for mass spectrometry purposes. The design of the source and its characteristics are presented. This paper comprises partly the PhD theses of J Ketelaer and C Smorra.

  6. Neutronics, Steady-State, and Transient Analyses for the Kazakhstan VVR-K Reactor with LEU Fuel

    SciTech Connect

    Hanan, N.A.; Garner, P.L.

    2015-01-01

    Calculations have been performed for steady state and postulated transients in the VVRK reactor at the Institute of Nuclear Physics (INP) in Alatau, Kazakhstan. These calculations have been performed at the request of staff of the INP who have performed similar calculations. Calculations were performed for the fresh low-enriched uranium (LEU) core and for four subsequent cores as beryllium is added as a radial reflector to maintain criticality during the first 15 cycles of operation. The calculations include neutronics parameters, steady-state power and temperature distributions, and response to transients. These calculations have been performed independently from those being performed by INP and serve as one step in the verification process.

  7. 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.

  8. 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.

  9. Steady-state inhibition model for the biodegradation of sulfonated amines in a packed bed reactor.

    PubMed

    Juárez-Ramírez, Cleotilde; Galíndez-Mayer, Juvencio; Ruiz-Ordaz, Nora; Ramos-Monroy, Oswaldo; Santoyo-Tepole, Fortunata; Poggi-Varaldo, Héctor

    2015-05-25

    Aromatic amines are important industrial products having in their molecular structure one or more aromatic rings. These are used as precursors for the synthesis of dyes, adhesives, pesticides, rubber, fertilizers and surfactants. The aromatic amines are common constituents of industrial effluents, generated mostly by the degradation of azo dyes. Several of them are a threat to human health because they can by toxic, allergenic, mutagenic or carcinogenic. The most common are benzenesulfonic amines, such as 4-ABS (4-aminobenzene sulfonic acid) and naphthalene sulfonic amines, such as 4-ANS (4-amino naphthalene sulfonic acid). Sometimes, the mixtures of toxic compounds are more toxic or inhibitory than the individual compounds, even for microorganisms capable of degrading them. Therefore, the aim of this study was to evaluate the degradation of the mixture 4-ANS plus 4-ABS by a bacterial community immobilized in fragments of volcanic stone, using a packed bed continuous reactor. In this reactor, the amines loading rates were varied from 5.5 up to 69 mg L(-1) h(-1). The removal of the amines was determined by high-performance liquid chromatography and chemical oxygen demand. With this information, we have studied the substrate inhibition of the removal rate of the aromatic amines during the degradation of the mixture of sulfonated aromatic amines by the immobilized microorganisms. Experimental results were fitted to parabolic, hyperbolic and linear inhibition models. The model that best characterizes the inhibition of the specific degradation rate in the biofilm reactor was a parabolic model with values of RXM=58.15±7.95 mg (10(9) cells h)(-1), Ks=0.73±0.31 mg L(-1), Sm=89.14±5.43 mg L(-1) and the exponent m=5. From the microbial community obtained, six cultivable bacterial strains were isolated and identified by sequencing their 16S rDNA genes. The strains belong to the genera Variovorax, Pseudomonas, Bacillus, Arthrobacter, Nocardioides and Microbacterium. This

  10. 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. Copyright © 2013 Elsevier Ltd. All rights reserved.

  11. Preliminary Experimental Results using a Steady State ICP Flow Reactor to Investigate Condensation Chemistry for Nuclear Forensics

    NASA Astrophysics Data System (ADS)

    Koroglu, Batikan; Armstrong, Mike; Cappelli, Mark; Chernov, Alex; Crowhurst, Jonathan; Mehl, Marco; Radousky, Harry; Rose, Timothy; Zaug, Joe

    2016-10-01

    The high temperature chemistry of rapidly condensing matter is under investigation using a steady state inductively coupled plasma (ICP) flow reactor. The objective is to study chemical processes on cooling time scales similar to that of a low yield nuclear fireball. The reactor has a nested set of gas flow rings that provide flexibility in the control of hydrodynamic conditions and mixing of chemical components. Initial tests were run using two different aqueous solutions (ferric nitrate and uranyl nitrate). Chemical reactants passing through the plasma torch undergo non-linear cooling from 10,000K to 1,000K on time scales of <0.1 to 0.5s depending on flow conditions. Optical spectroscopy measurements were taken at different positions along the flow axis to observe the in situ spatial and temporal evolution of chemical species at different temperatures. The current data offer insights into the changes in oxide chemistry as a function of oxygen fugacity. The time resolved measurements will also serve as a validation target for the development of kinetic models that will be used to describe chemical fractionation during nuclear fireball condensation. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

  12. 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.

  13. 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

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. Fast Reactors

    NASA Astrophysics Data System (ADS)

    Esposito, S.; Pisanti, O.

    The following sections are included: * Elementary Considerations * The Integral Equation to the Neutron Distribution * The Critical Size for a Fast Reactor * Supercritical Reactors * Problems and Exercises

  19. 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

  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. 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.

  2. 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.

  3. Radioactivity computation of steady-state and pulsed fusion reactors operation

    SciTech Connect

    Attaya, H.

    1994-06-01

    Different mathematical methods are used to calculate the nuclear transmutation in steady-state and pulsed neutron irradiation. These methods are the Schuer decomposition, the eigenvector decomposition, and the Pade approximation of the matrix exponential function. In the case of the linear decay chain approximation, a simple algorithm is used to evaluate the transition matrices.

  4. 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.

  5. 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... governmental bodies, and Federally- recognized Indian tribes do not need to address the standing requirements... on electronic storage media. Participants may not submit paper copies of their filings unless...

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. The bacterial community inside a fluidized bed reactor, receiving mine-affected water converges to an overall steady state structure correlating with better reactor performance

    NASA Astrophysics Data System (ADS)

    Hatam, I.; Baldwin, S. A.

    2016-12-01

    Water seepage from mine sites often contains dissolved metalloids such as oxidized forms of Se, in concentrations above regulated limits for fresh water. Discharge of these selenium contaminated waters to the environment can have serious repercussions due to the propensity of oxidized Se to bioaccumulate with consequent toxicity to wildlife. Therefore treatment is necessary prior to water discharge. One approach for this is using bioreactors which rely on bacterial metabolism to reduce oxidized Se into less mobile forms. Often these bioreactors are inoculated with one or two species of bacteria, charged with the task of reducing oxidized selenium. This strategy can be problematic since bacteria found naturally in the water (and soil) often compete with the starter culture for resources and niche space. This also assumes other terminal electron acceptors such as NO3, often found where explosives are utilized in the mining process, will not be preferentially used by the starter culture. Here we evaluated the performance of a fluidized bed reactor designed to treat mine affected ground water contaminated with oxidized Se and NO3, in the context of the development of its bacteria community. The reactor was inoculated with one species of bacteria and we examined the changes to the composition of the bacterial population over the duration of 7 months. Our results show that as early as four weeks from the beginning of operations the community was comprised of >90 OTUs belonging to 13 bacterial classes. Over time, the community became less variable in composition and this convergence was associated with higher percent of NO3 and Se removal. Thus, our results suggest that while bacteria from the environment do occupy niches within the bioreactor, the in-situ conditions select for environmental taxa capable of performing contaminants removal better than the starter culture.

  14. Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

    DOE PAGES

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; ...

    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.

  15. 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.

  16. 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.

  17. Effect of small-scale biomass gasification at the state of refractory lining the fixed bed reactor

    SciTech Connect

    Janša, Jan Peer, Vaclav Pavloková, Petra

    2016-06-30

    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.

  18. Detector of the reactor AntiNeutrino based on Solid-state plastic Scintillator (DANSS). Status and first results.

    NASA Astrophysics Data System (ADS)

    Alekseev, I.; Belov, V.; Brudanin, V.; Danilov, M.; Egorov, V.; Filosofov, D.; Fomina, M.; Hons, Z.; Kazartsev, S.; Kobyakin, A.; Kuznetsov, A.; Machikhiliyan, I.; Medvedev, D.; Nesterov, V.; Olshevsky, A.; Ponomarev, D.; Rozova, I.; Rumyantseva, N.; Rusinov, V.; Salamatin, A.; Shevchik, Ye; Shirchenko, M.; Shitov, Yu; Skrobova, N.; Starostin, A.; Svirida, D.; Tarkovsky, E.; Tikhomirov, I.; Vlášek, J.; Zhitnikov, I.; Zinatulina, D.

    2017-01-01

    A detector of the reactor antineutrino based on a cubic meter of plastic scintillator is installed below 3.1 GW industrial reactor. The detector is placed on a movable platform which allows to change the distance to the reactor core center in the range 10.7-12.7 m. 2500 scintillator strips are read out individually by SiPMs and in groups of 50 by PMTs. In addition to the overburden by the reactor (50 m w.e.) the detector has multilayer passive shielding and active muon veto. Inverse beta-decay count rate of about 5000 events per day in the fiducial volume (78% of the detector) with about 5% of cosmic background has been reached. DANSS is sensitive to sterile neutrino in the most interesting region of mixing parameter space. The article covers the detector status and performance, as well as the first results.

  19. 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)

  20. 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.

  1. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1961-09-01

    A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. Retrofit Russian research reactors

    SciTech Connect

    Mabe, W.

    1993-04-01

    A likely source for enriched uranium for production of a gun-type bomb might be a research reactor. A state or terrorist organization would find the technical process for separating uranium from the reactor fuel plates is simple and well-published. An unguarded research reactor could be found in the former Soviet Union. Russia and the former republics have seen an increasing number of terrorist incidents, including hijackings and bombings. Recognizing the danger, Russia and the U.S. have explored means of safeguarding former Soviet weapons materials. This article describes some of the plans to reduce the risk of nuclear materials being obtained for illicit weapons production.

  7. 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.

  8. 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.

  9. 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.

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

    SciTech Connect

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

    2013-07-01

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

  11. K-Reactor readiness

    SciTech Connect

    Rice, P.D.

    1991-12-04

    This document describes some of the more significant accomplishments in the reactor restart program and details the magnitude and extent of the work completed to bring K-Reactor to a state of restart readiness. The discussion of restart achievements is organized into the three major categories of personnel, programs, and plant. Also presented is information on the scope and extent of internal and external oversight of the efforts, as well as some details on the startup plan.

  12. K-Reactor readiness

    SciTech Connect

    Rice, P.D.

    1991-12-04

    This document describes some of the more significant accomplishments in the reactor restart program and details the magnitude and extent of the work completed to bring K-Reactor to a state of restart readiness. The discussion of restart achievements is organized into the three major categories of personnel, programs, and plant. Also presented is information on the scope and extent of internal and external oversight of the efforts, as well as some details on the startup plan.

  13. Long-term ammonia removal in a coconut fiber-packed biofilter: analysis of N fractionation and reactor performance under steady-state and transient conditions.

    PubMed

    Baquerizo, Guillermo; Maestre, Juan P; Machado, Vinicius C; Gamisans, Xavier; Gabriel, David

    2009-05-01

    A comprehensive study of long-term ammonia removal in a biofilter packed with coconut fiber is presented under both steady-state and transient conditions. Low and high ammonia loads were applied to the reactor by varying the inlet ammonia concentration from 90 to 260 ppm(v) and gas contact times ranging from 20 to 36 s. Gas samples and leachate measurements were periodically analyzed and used for characterizing biofilter performance in terms of removal efficiency (RE) and elimination capacity (EC). Also, N fractions in the leachate were quantified to both identify the experimental rates of nitritation and nitratation and to determine the N leachate distribution. Results showed stratification in the biofilter activity and, thus, most of the NH(3) removal was performed in the lower part of the reactor. An average EC of 0.5 kg N-NH(3)m(-3)d(-1) was obtained for the whole reactor with a maximum local average EC of 1.7 kg N-NH(3)m(-3)d(-1). Leachate analyses showed that a ratio of 1:1 of ammonium and nitrate ions in the leachate was obtained throughout steady-state operation at low ammonia loads with similar values for nitritation and nitratation rates. Low nitratation rates during high ammonia load periods occurred because large amounts of ammonium and nitrite accumulated in the packed bed, thus causing inhibition episodes on nitrite-oxidizing bacteria due to free ammonia accumulation. Mass balances showed that 50% of the ammonia fed to the reactor was oxidized to either nitrite or nitrate and the rest was recovered as ammonium indicating that sorption processes play a fundamental role in the treatment of ammonia by biofiltration.

  14. 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

  15. 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)

  16. 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. Copyright © 2012 Elsevier Ltd. All rights reserved.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  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. Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

    NASA Astrophysics Data System (ADS)

    Zhang, Da-Lin; Qiu, Sui-Zheng; Liu, Chang-Liang; Su, Guang-Hui

    2008-08-01

    The Molten Salt Reactor (MSR), one of the `Generation IV' concepts, is a liquid-fuel reactor, which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt. The study on its neutronics considering the fuel salt flow, which is the base of the thermal-hydraulic calculation and safety analysis, must be done. In this paper, the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method, and the discretization equations are computed by the source iteration method. The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained. The numerical calculated results show that, the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially the long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition. Supported by National Nature Science Foundation of China (10575079)

  4. 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.

  5. Reactor building

    SciTech Connect

    Hista, J. C.

    1984-09-18

    Reactor building comprising a vessel shaft anchored in a slab which is peripherally locked. This reactor building comprises a confinement enclosure within which are positioned internal structures constituted by an internal structure floor, a vessel shaft, a slab being positioned between the general floor and the internal structure floor, the vesse

  6. 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.

  7. 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).

  8. F Reactor Inspection

    ScienceCinema

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2016-07-12

    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."

  9. 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."

  10. 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.

  11. 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.

  12. 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

  13. 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.

  14. NEUTRONIC REACTORS

    DOEpatents

    Wigner, E.P.

    1960-11-22

    A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

  15. 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.

  16. Reactor apparatus

    DOEpatents

    Echtler, J. Paul

    1981-01-01

    A reactor apparatus for hydrocracking a polynuclear aromatic hydrocarbonaceous feedstock to produce lighter hydrocarbon fuels by contacting the hydrocarbonaceous feedstock with hydrogen in the presence of a molten metal halide catalyst.

  17. 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.

  18. Reactor Engineering

    NASA Astrophysics Data System (ADS)

    Lema, Juan M.; López, Carmen; Eibes, Gemma; Taboada-Puig, Roberto; Moreira, M. Teresa; Feijoo, Gumersindo

    In this chapter, the engineering aspects of processes catalyzed by peroxidases will be presented. In particular, a discussion of the existing technologies that utilize peroxidases for different purposes, such as the removal of recalcitrant compounds or the synthesis of polymers, is analyzed. In the first section, the essential variables controlling the process will be investigated, not only those that are common in any enzymatic system but also those specific to peroxidative reactions. Next, different reactor configurations and operational modes will be proposed, emphasizing their suitability and unsuitability for different systems. Finally, two specific reactors will be described in detail: enzymatic membrane reactors and biphasic reactors. These configurations are especially valuable for the treatment of xenobiotics with high and poor water solubility, respectively.

  19. 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.

  20. 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)

  1. 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.

  2. NEUTRONIC REACTORS

    DOEpatents

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  3. 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)

  4. 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.

  5. 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.

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

    SciTech Connect

    Fourmentel, D.; Radulovic, V.; Barbot, L.; Villard, J-F.; Zerovnik, G.; Snoj, L.; Tarchalski, M.; Pytel, K.; Malouch, F.

    2015-07-01

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

  7. The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

    SciTech Connect

    Dunn, Floyd E.; Hu, Lin-wen; Wilson, Erik

    2016-12-01

    The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.

  8. General Criteria and Operation Limits of a Steady-State Fusion Reactor with Respect to Plasma-Material Interaction

    NASA Astrophysics Data System (ADS)

    Naujoks, D.

    2010-05-01

    The magnetic confinement of a hot plasma is the most promising concept to realize controlled thermonuclear fusion on earth. In the last years of intense research activities in the frame of broad international collaboration, it became clear that on the way to a stationary operating fusion reactor not only questions of plasma heating, transport and stability but also the problems associated with the choice of plasma facing materials are decisive. These issues cannot be decoupled from each other [1]. It is demonstrated that both sides, the plasma and the wall, exhibit mutual dependences. Burning conditions will not be achieved without careful adaptation of the chosen materials to the developed plasma scenarios and vice versa. Integrated concepts are required.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 76 FR 43733 - The University of Utah; Facility Operating License No. R-126; Notice of Acceptance for Docketing...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-07-21

    ..., Research and Test Reactors Licensing Branch, Division of Policy and Rulemaking, Office of Nuclear Reactor... Renewal for Research Reactors'' (ISG) which can be obtained from the Commission's public document room... licensee) to operate the University of Utah TRIGA Reactor (UUTR) at a maximum steady-state thermal power...

  15. 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.

  16. 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

    2013-10-07

    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.

  17. Reactor performance and microbial community dynamics during anaerobic co-digestion of municipal wastewater sludge with restaurant grease waste at steady state and overloading stages.

    PubMed

    Razaviarani, Vahid; Buchanan, Ian D

    2014-11-01

    Linkage between reactor performance and microbial community dynamics was investigated during mesophilic anaerobic co-digestion of restaurant grease waste (GTW) with municipal wastewater sludge (MWS) using 10L completely mixed reactors and a 20day SRT. Test reactors received a mixture of GTW and MWS while control reactors received only MWS. Addition of GTW to the test reactors enhanced the biogas production and methane yield by up to 65% and 120%, respectively. Pyrosequencing revealed that Methanosaeta and Methanomicrobium were the dominant acetoclastic and hydrogenotrophic methanogen genera, respectively, during stable reactor operation. The number of Methanosarcina and Methanomicrobium sequences increased and that of Methanosaeta declined when the proportion of GTW in the feed was increased to cause an overload condition. Under this overload condition, the pH, alkalinity and methane production decreased and VFA concentrations increased dramatically. Candidatus cloacamonas, affiliated within phylum Spirochaetes, were the dominant bacterial genus at all reactor loadings.

  18. 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.

  19. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.

    1957-10-01

    A reactor of the type which preferably uses plutonium as the fuel and a liquid moderator, preferably ordinary water, and which produces steam within the reactor core due to the heat of the chain reaction is described. In the reactor shown the fuel elements are essentially in the form of trays and are ventically stacked in spaced relationship. The water moderator is continuously supplied to the trays to maintain a constant level on the upper surfaces of the fuel element as it is continually evaporated by the heat. The steam passes out through the spaces between the fuel elements and is drawn off at the top of the core. The fuel elements are clad in aluminum to prevent deterioration thereof with consequent contamimation of the water.

  20. 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)

  1. 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.

  2. 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.

  3. Catalytic reactor

    SciTech Connect

    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.

  4. 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.

  5. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Szilard, L.

    1957-09-24

    Reactors of the type employing plates of natural uranium in a moderator are discussed wherein the plates are um-formly disposed in parallel relationship to each other thereby separating the moderator material into distinct and individual layers. Each plate has an uninterrupted sunface area substantially equal to the cross-sectional area of the active portion of the reactor, the particular size of the plates and the volume ratio of moderator to uranium required to sustain a chain reaction being determinable from the known purity of these materials and other characteristics such as the predictable neutron losses due to the formation of radioactive elements of extremely high neutron capture cross section.

  6. 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.

  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. 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.

  9. 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.

  10. 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.

  11. NEUTRONIC REACTORS

    DOEpatents

    Anderson, H.L.

    1958-10-01

    The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

  12. 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.

  13. 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.

  14. 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.

  15. 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.}

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Concept for LEU Burst Reactor

    SciTech Connect

    Klein, Steven Karl; Kimpland, Robert Herbert

    2016-03-07

    Design and performance of a proposed LEU burst reactor are sketched. Salient conclusions reached are the following: size would be ~1,500 kg or greater, depending on the size of the central cavity; internal stresses during burst require split rings for relief; the reactor would likely require multiple control and safety rods for fine control; the energy spectrum would be comparable to that of HEU machines; and burst yields and steady-state power levels will be significantly greater in an LEU reactor.

  1. 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)

  2. Reactor antineutrinos and nuclear physics

    NASA Astrophysics Data System (ADS)

    Balantekin, A. B.

    2016-11-01

    Short-baseline reactor neutrino experiments successfully measured the neutrino parameters they set out to measure, but they also identified a shape distortion in the 5-7 MeV range as well as a reduction from the predicted value of the flux. Nuclear physics input into the calculations of reactor antineutrino spectra needs to be better refined if this anomaly is to be interpreted as due to sterile neutrino states.

  3. FIBWR: a steady-state core flow distribution code for boiling water reactors code verification and qualification report. Final report

    SciTech Connect

    Ansari, A.F.; Gay, R.R.; Gitnick, B.J.

    1981-07-01

    A steady-state core flow distribution code (FIBWR) is described. The ability of the recommended models to predict various pressure drop components and void distribution is shown by comparison to the experimental data. Application of the FIBWR code to the Vermont Yankee Nuclear Power Station is shown by comparison to the plant measured data.

  4. 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.

  5. 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.

  6. 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/.

  7. 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.

  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. Neutronic reactor

    DOEpatents

    Lewis, Warren R.

    1978-05-30

    A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

  11. 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.

  12. 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.

  13. 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)

  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. Space reactors

    NASA Astrophysics Data System (ADS)

    Ranken, W. A.

    1983-01-01

    Progress in design studies and technology for the SP-100 Project - successor to the Space Power Advanced Reactor (SPAR) Project - is reported for the period October 1, 1981 to March 31, 1982. The basis for selecting a high-temperature, UO2-fueled, heat-pipe-cooled reactor with a thermoelectric conversion system as the 100/kW-sub e/ reference design has been reviewed. Although no change has been made in the general concept, design studies have been done to investigate various reactor/conversion system coupling methods and core design modifications. Thermal and mechanical finite element modeling and three dimensional Monte Carlo analysis of a core with individual finned fuel elements are reported. Studies of unrestrained fuel irradiation data are discussed that are relevant both to the core modeling work and to the design and fabrication of the first in-pile irradiation test, which is also reported. Work on lithium-filled core heat pipe development is described, including the attainment of 15.6 kW/sub t/ operation at 1525 K for a 2-m-long heat pipe with a 15.7-mm outside diameter. The successful operation of a 5.5-m-long, lightweight potassium/titanium heat pipe at 760 K is described, and test results of a thermoelectric module with GaP-modified SiGe thermoelectric elements are presented.

  16. A novel plant protection strategy for transient reactors

    NASA Astrophysics Data System (ADS)

    Bhattacharyya, Samit K.; Lipinski, Walter C.; Hanan, Nelson A.

    A novel plant protection system designed for use in the TREAT Upgrade (TU) reactor is described. The TU reactor is designed for controlled transient operation in the testing of reactor fuel behavior under simulated reactor accident conditions. Safe operation of the reactor is of paramount importance and the Plant Protection System (PPS) had to be designed to exacting requirements. Researchers believe that the strategy developed for the TU has potential application to the multimegawatt space reactors and represents the state of the art in terrestrial transient reactor protection systems.

  17. 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.

  18. 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.)

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. Photocatalytic reactor

    DOEpatents

    Bischoff, B.L.; Fain, D.E.; Stockdale, J.A.D.

    1999-01-19

    A photocatalytic reactor is described for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane. 4 figs.

  4. [A new correction method for radionuclide formation in neutron activation analysis using a reactor power meter coupled with a microcomputer].

    PubMed

    Hirai, S; Yoshino, Y; Suzuki, S; Horiuchi, N

    1982-05-01

    Neutron flux and irradiation time should be accurately known in neutron activation analysis using very short lived nuclides in which conventional monitoring methods i.e., a comparator method, flux monitor method and so on cannot be used satisfactorily. Especially, fluctuation of neutron flux has not been corrected. We noted a change of reactor power at a pneumatic operation, and found out a new correction method for its correction in activation analysis. In our small nuclear reactor, TRIGA-II, the reactor power increased rapidly a few % when a pneumatic-operated capsule arrived at a core of the reactor, and decreased when the capsule left from the core. If the duration between these two changes of the reactor power is equal to the irradiation time, and that the reactor power is proportional to the neutron flux, we can regard an activity formation as a time integration of the reactor power. Then, the correction system was made of a reactor power meter, a V-F converter (voltage to frequency converter), a clock time, a counter, a microcomputer, electric circuits and so on. The signal of the reactor power during the irradiation was counted through the V-F converter, and was accumulated in a memory of the microcomputer. The neutron fluence was calculated in this microcomputer. This method was examined by means of activation of copper and selenium standard samples by 9-11 sec irradiations. The observed activity involved +/- 10% error. However, the error in the corrected activity was decreased to a few % using this correction method. As a result, we found that this method can be used to obtain accurate value for radionuclide formation.

  5. 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.

  6. Assessment of torsatrons as reactors

    SciTech Connect

    Lyon, J.F. ); Painter, S.L. )

    1992-12-01

    Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R[sub 0] = 6.6-8.8 m, on-axis magnetic field B[sup 0] = 4.8-7.5 T, B[sub max] (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions.

  7. Small Reactor for Deep Space Exploration

    ScienceCinema

    none,

    2016-07-12

    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.

  8. Canadian reactor delays add to agency's troubles

    NASA Astrophysics Data System (ADS)

    Howell, Elizabeth

    2009-08-01

    A plan to replace two aging nuclear reactors in Ontario, Canada, with Advanced CANDU Reactors (ACR) has been rejected by the provincial government due to concerns over ballooning costs. The rejected bid, which was made by the state-owned Atomic Energy of Canada Limited (AECL), adds to the woes of an agency that is already facing rumoured privatization of its business to build and maintain nuclear reactors.

  9. D-D tokamak reactor studies

    SciTech Connect

    Evans, K.E. Jr.; Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Finn, P.A.; Jung, J.; Mattas, R.F.; Misra, B.; Smith, D.L.; Stevens, H.C.

    1980-11-01

    A tokamak D-D reactor design, utilizing the advantages of a deuterium-fueled reactor but with parameters not unnecessarily extended from existing D-T designs, is presented. Studies leading to the choice of a design and initial studies of the design are described. The studies are in the areas of plasma engineering, first-wall/blanket/shield design, magnet design, and tritium/fuel/vacuum requirements. Conclusions concerning D-D tokamak reactors are stated.

  10. 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.

  11. 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.

  12. Hybrid adsorptive membrane reactor

    DOEpatents

    Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

    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.

  13. Performance of membrane fixed biocatalyst reactors. I: Membrane reactor systems and modelling.

    PubMed

    Prenosil, J E; Hediger, T

    1988-06-05

    Enzymatic membrane reactors are discussed according to the state of biocatalyst and driving force of reaction. Particular attention is given to the Capillary Membrane Fixed Enzyme Reactor (CAMFER) for its favorable characteristics. It is shown that, for a practical range of operation conditions, both kinetic and mass transfer effects must be considered simultaneously. Three modes of operation were investigated in detail using enzymatic lactose hydrolysis as a model reaction: Diffusional reactor, Recycle reactor, and Backflush reactor. In the comparison, superior performance of the CAMFER in diffusional mode was clearly demonstrated.

  14. 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.

  15. 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.

  16. 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.

  17. 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)

  18. NEUTRONIC REACTOR

    DOEpatents

    McGarry, R.J.

    1958-04-22

    Fluid-cooled nuclear reactors of the type that utilize finned uranium fuel elements disposed in coolant channels in a moderater are described. The coolant channels are provided with removable bushings composed of a non- fissionable material. The interior walls of the bushings have a plurality of spaced, longtudinal ribs separated by grooves which receive the fins on the fuel elements. The lands between the grooves are spaced from the fuel elements to form flow passages, and the size of the now passages progressively decreases as the dlstance from the center of the core increases for the purpose of producing a greater cooling effect at the center to maintain a uniform temperature throughout the core.

  19. Benchmark Evaluation of the Neutron Radiography (NRAD) Reactor Upgraded LEU-Fueled Core

    SciTech Connect

    John D. Bess

    2001-09-01

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. The final upgraded core configuration with 64 fuel elements has been completed. Evaluated benchmark measurement data include criticality, control-rod worth measurements, shutdown margin, and excess reactivity. Dominant uncertainties in keff include the manganese content and impurities contained within the stainless steel cladding of the fuel and the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 nuclear data are approximately 1.4% greater than the benchmark model eigenvalue, supporting contemporary research regarding errors in the cross section data necessary to simulate TRIGA-type reactors. Uncertainties in reactivity effects measurements are estimated to be ~10% with calculations in agreement with benchmark experiment values within 2s. The completed benchmark evaluation de-tails are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Experiments (IRPhEP Handbook). Evaluation of the NRAD LEU cores containing 56, 60, and 62 fuel elements have also been completed, including analysis of their respective reactivity effects measurements; they are also available in the IRPhEP Handbook but will not be included in this summary paper.

  20. Test of a prototype neutron spectrometer based on diamond detectors in a fast reactor

    SciTech Connect

    Osipenko, M.; Ripani, M.; Ricco, G.; Caiffi, B.; Pompili, F.; Pillon, M.; Angelone, M.; Verona-Rinati, G.; Cardarelli, R.; Argiro, S.

    2015-07-01

    A prototype of neutron spectrometer based on diamond detectors has been developed. This prototype consists of a {sup 6}Li neutron converter sandwiched between two CVD diamond crystals. The radiation hardness of the diamond crystals makes it suitable for applications in low power research reactors, while a low sensitivity to gamma rays and low leakage current of the detector permit to reach good energy resolution. A fast coincidence between two crystals is used to reject background. The detector was read out using two different electronic chains connected to it by a few meters of cable. The first chain was based on conventional charge-sensitive amplifiers, the other used a custom fast charge amplifier developed for this purpose. The prototype has been tested at various neutron sources and showed its practicability. In particular, the detector was calibrated in a TRIGA thermal reactor (LENA laboratory, University of Pavia) with neutron fluxes of 10{sup 8} n/cm{sup 2}s and at the 3 MeV D-D monochromatic neutron source named FNG (ENEA, Rome) with neutron fluxes of 10{sup 6} n/cm{sup 2}s. The neutron spectrum measurement was performed at the TAPIRO fast research reactor (ENEA, Casaccia) with fluxes of 10{sup 9} n/cm{sup 2}s. The obtained spectra were compared to Monte Carlo simulations, modeling detector response with MCNP and Geant4. (authors)

  1. Genotoxicity of neutrons in Drosophila melanogaster. Somatic mutation and recombination induced by reactor neutrons.

    PubMed

    Guzmán-Rincón, J; Delfín-Loya, A; Ureña-Núñez, F; Paredes, L C; Zambrano-Achirica, F; Graf, U

    2005-08-01

    This paper describes the observation of a direct relationship between the absorbed doses of neutrons and the frequencies of somatic mutation and recombination using the wing somatic mutation and recombination test (SMART) of Drosophila melanogaster. This test was used for evaluating the biological effects induced by neutrons from the Triga Mark III reactor of Mexico. Two different reactor power levels were used, 300 and 1000 kW, and two absorbed doses were tested for each power level: 1.6 and 3.2 Gy for 300 kW and 0.84 and 1.7 Gy for 1000 kW. A linear relationship was observed between the absorbed dose and the somatic mutation and recombination frequencies. Furthermore, these frequencies were dependent on larval age: In 96-h-old larvae, the frequencies were increased considerably but the sizes of the spots were smaller than in 72-h-old larvae. The analysis of the balancer-heterozygous progeny showed a linear absorbed dose- response relationship, although the responses were clearly lower than found in the marker-trans-heterozygous flies. Approximately 65% of the genotoxicity observed is due to recombinational events. The results of the study indicate that thermal and fast neutrons are both mutagenic and recombinagenic in the D. melanogaster wing SMART, and that the frequencies are dependent on neutron dose, reactor power, and the age of the treated larvae.

  2. Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

    SciTech Connect

    Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.; Marshall, Margaret A.

    2014-11-04

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental keff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  3. Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

    DOE PAGES

    Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.; ...

    2014-11-04

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental keff come from uncertainties in the manganese content and impurities in the stainless steel fuel cladding asmore » well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

  4. Determination of Long-Lived Neutron Activation Products in Reactor Shielding Concrete Samples

    SciTech Connect

    Zagar, Tomaz; Ravnik, Matjaz

    2002-10-15

    The results of activation studies of TRIGA research reactor concrete shielding are given. Samples made of ordinary and barytes concrete were irradiated in the reactor to simulate neutron activation in the shielding concrete. Long-lived neutron-induced gamma-ray-emitting radioactive nuclides were measured in the samples with a high-purity germanium detector. The most active long-lived radioactive nuclides in the ordinary concrete samples were found to be {sup 60}Co and {sup 152}Eu. In the barytes concrete samples, the most active long-lived radioactive nuclides were {sup 60}Co, {sup 133}Ba, and {sup 152}Eu. Activation in the concrete was also calculated using the ORIGEN2 code and compared to experimental results. Simple radioactive nuclide generation and depletion calculation using one-group cross-section libraries provided together with the ORIGEN2 code did not give conservative results. Significant discrepancies were observed for some nuclides. For accurate long-lived radioactive nuclide generation in reactor shielding, material-specific cross-section libraries should be generated and verified by measurement.

  5. Nuclear reactor neutron shielding

    DOEpatents

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  6. 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.

  7. 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.

  8. Evaluation of Torsatrons as reactors

    SciTech Connect

    Lyon, J.F.; Gulec, K.; Miller, R.L.; El-Guebaly, L.

    1994-03-01

    Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors. This scoping study, which uses an integrated cost-minimization code that incorporates costing and reactor component models self-consistently with a 1-D energy transport calculation, shows that a torsatron reactor could also be economically competitive with a tokamak reactor. The projected cost of electricity (COE) estimated using the Advanced Reactor Innovation and Evaluation Studies (ARIES) costing algorithms is 65.6 mill/kW(e)h in constant 1992 dollars for a reference 1-GW(e) Compact Torsatron reactor case. The COE is relatively insensitive (<10% variation) over a wide range of assumptions, including variations in the maximum field allowed on the coils, the coil elongation, the shape of the density profile, the beta limit, the confinement multiplier, and the presence of a large loss region for alpha particles. The largest variations in the COE occur for variations in the electrical power output demanded and the plasma-coil separation ratio.

  9. Plasma instrumentation for fusion power reactor control

    SciTech Connect

    Sager, G.T.; Bauer, J.F.; Maya, I.; Miley, G.H.

    1985-07-01

    Feedback control will be implemented in fusion power reactors to guard against unpredicted behavior of the plant and to assure desirable operation. In this study, plasma state feedback requirements for plasma control by systems strongly coupled to the plasma (magnet sets, RF, and neutral beam heating systems, and refueling systems) are estimated. Generic considerations regarding the impact of the power reactor environment on plasma instrumentation are outlined. Solutions are proposed to minimize the impact of the power reactor environment on plasma instrumentation. Key plasma diagnostics are evaluated with respect to their potential for upgrade and implementation as power reactor instruments.

  10. 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.

  11. 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.

  12. Preliminary study of degradation from neutron effects of core-structural materials of Thai Research Reactor TRR-1/M1

    NASA Astrophysics Data System (ADS)

    Ampornrat, P.; Boonsuwan, P.; Sangkaew, S.; Angwongtrakool, T.

    2017-06-01

    Thai research reactor went first critical in 1962. The reactor was converted in 1977 from an MTR-type with high-enriched uranium fuel to a TRIGA-MARK III type using low-enriched uranium fuel, called TRR-1/M1. Since the TRR-1/M1 has been operated for almost 40 years, degradation of reactor structural materials is expected. In this preliminary study, the potential degradation from neutron effects of core-structural materials, e.g., fuel clad (SS304) and core components (Al6061) were studied. Assessment included calculation of neutron energy, flux and fluence in the reactor core to evaluate displacement rate (dpa) and irradiation effects on the material properties. Results showed maximum displacement rates on SS304 was 5.24×10-8 per cm3·sec and on Al6061 was 1.14×10-8 per cm3·sec. The corresponding maximum displacement levels were ∼17 dpa for SS304, and ∼4 dpa for Al6061. At these levels of displacement, it is possible for the materials to result in tensile strength increasing and ductility reduction. Further inspection on the core-structural materials needs to be conducted to validate the assessment results from this study.

  13. Neutronics and steady-state thermal hydraulics analysis for the HEU, mixed HEU-LEU and the first full LEU cores of WWR-SM reactor at INP AS RUZ

    SciTech Connect

    Baytelesov, S.A.; Dosimbaev, A.A.; Kungurov, F.R.; Salikhbaev, U.S.

    2008-07-15

    The WWR-SM reactor at the Institute of Nuclear Physics of the Academy of Sciences (INP AS) in Uzbekistan is preparing for the conversion from HEU (36%) fuel to LEU (19.8%) fuel. During this conversion, the HEU fuel assemblies (IRT-3M FA) being discharged at the end of each cycle will be replaced by LEU fuel assemblies (IRT-4M FA); this gradual conversion requires 9 cycles. Conversion to LEU without loss of performance for the present experimental program requires the size of the core to increase from 18 to 20 fuel assemblies and the power of the reactor to increase from 10 to 11 MW. The safety analysis report for this conversion process has been prepared. This paper presents the methods and results for the neutronics analysis (burnup, power distributions and shutdown margin), the steady-state thermal hydraulics analysis and the kinetics parameters for the HEU, all mixed and the first full LEU cores. (author)

  14. 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.

  15. Verification and Validation of Neutronic/Thermalhydraulic 3D-Time Dependent Model for Treatment of Super-critical States of Light water Research Reactors Accidents

    SciTech Connect

    Khaled, S.M.

    2015-07-01

    This work presents the Verification and testing both the neutronic and thermal-hydraulics response of the positive reactivity-initiated power excursion accidents in small light water research reactors. Some research reactors have to build its own severe accidents code system. In this sense, a 3D space-time-dependent neutron diffusion models with thermal hydraulic feedback have been introduced, compared and tested both experimentally at criticality 14-cent and theoretically up to 1.5 $ with a number of similar codes. The results shows that no expected core failure or moderator boiling. (author)

  16. Update on reactors and reactor instruments in Asia

    NASA Astrophysics Data System (ADS)

    Rao, K. R.

    1991-10-01

    The 1980s have seen the commissioning of several medium flux (∼10 14 neutrons/cm 2s) research reactors in Asia. The reactors are based on indigenous design and development in India and China. At Dhruva reactor (India), a variety of neutron spectrometers have been established that have provided useful data related to the structure of high- Tc materials, phonon density of states, magnetic moment distributions and micellar aggregation during the last couple of years. Polarised neutron analysis, neutron interferometry and neutron spin echo methods are some of the new techniques under development. The spectrometers and associated automaton, detectors and neutron guides have all been indigenously developed. This paper summarises the developments and on-going activities in Bangladesh, China, India, Indonesia, Korea, Malaysia, the Philippines and Thailand.

  17. 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.

  18. 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

  19. 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.

  20. 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.

  1. 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

  2. Accelerator based fusion reactor

    NASA Astrophysics Data System (ADS)

    Liu, Keh-Fei; Chao, Alexander Wu

    2017-08-01

    A feasibility study of fusion reactors based on accelerators is carried out. We consider a novel scheme where a beam from the accelerator hits the target plasma on the resonance of the fusion reaction and establish characteristic criteria for a workable reactor. We consider the reactions d+t\\to n+α,d+{{}3}{{H}\\text{e}}\\to p+α , and p+{{}11}B\\to 3α in this study. The critical temperature of the plasma is determined from overcoming the stopping power of the beam with the fusion energy gain. The needed plasma lifetime is determined from the width of the resonance, the beam velocity and the plasma density. We estimate the critical beam flux by balancing the energy of fusion production against the plasma thermo-energy and the loss due to stopping power for the case of an inert plasma. The product of critical flux and plasma lifetime is independent of plasma density and has a weak dependence on temperature. Even though the critical temperatures for these reactions are lower than those for the thermonuclear reactors, the critical flux is in the range of {{10}22}-{{10}24}~\\text{c}{{\\text{m}}-2}~{{\\text{s}}-1} for the plasma density {ρt}={{10}15}~\\text{c}{{\\text{m}}-3} in the case of an inert plasma. Several approaches to control the growth of the two-stream instability are discussed. We have also considered several scenarios for practical implementation which will require further studies. Finally, we consider the case where the injected beam at the resonance energy maintains the plasma temperature and prolongs its lifetime to reach a steady state. The equations for power balance and particle number conservation are given for this case.

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

    SciTech Connect

    Parma, Edward J., Jr.

    2009-06-01

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

  4. Comparing the new generation accelerator driven subcritical reactor system (ADS) to traditional critical reactors

    NASA Astrophysics Data System (ADS)

    Kemah, Elif; Akkaya, Recep; Tokgöz, Seyit Rıza

    2017-02-01

    In recent years, the accelerator driven subcritical reactors have taken great interest worldwide. The Accelerator Driven System (ADS) has been used to produce neutron in subcritical state by the external proton beam source. These reactors, which are hybrid systems, are important in production of clean and safe energy and conversion of radioactive waste. The ADS with the selection of reliability and robust target materials have been the new generation of fission reactors. In addition, in the ADS Reactors the problems of long-lived radioactive fission products and waste actinides encountered in the fission process of the reactor during incineration can be solved, and ADS has come to the forefront of thorium as fuel for the reactors.

  5. 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.

  6. NCSU Reactor Sharing Program. Final technical report, [September 1, 1980--August 29, 1992

    SciTech Connect

    Perez, P.B.

    1993-11-10

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities.

  7. 75 FR 68629 - Massachusetts Institute of Technology Reactor Notice of Issuance of Renewed Facility Operating...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-08

    ...-reflected research reactor licensed to operate at a steady-state power level of 6 megawatts thermal power..., Research and Test Reactors Licensing Branch, Division of Policy and Rulemaking, Office of Nuclear Reactor... COMMISSION Massachusetts Institute of Technology Reactor Notice of Issuance of Renewed Facility...

  8. Hybrid plasmachemical reactor

    SciTech Connect

    Lelevkin, V. M. Smirnova, Yu. G.; Tokarev, A. V.

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  9. Attrition reactor system

    SciTech Connect

    Scott, Charles D.; Davison, Brian H.

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  10. Attrition reactor system

    SciTech Connect

    Scott, C.D.; Davison, B.H.

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  11. 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.

  12. 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.

  13. Analysis of the Reduction Rate of Hematite Concentrate Particles in the Solid State by H2 or CO in a Drop-Tube Reactor Through CFD Modeling

    NASA Astrophysics Data System (ADS)

    Fan, De-Qiu; Sohn, H. Y.; Elzohiery, Mohamed

    2017-10-01

    The kinetic analysis of the reduction of hematite concentrate particles by individual reducing gas H2 or CO was performed using a computational fluid dynamics (CFD)-based approach in this paper. The particle residence time was calculated through the integration of the equation of particle motion. Non-uniform particle temperature profiles inside the reactor were obtained, and were taken into consideration for the kinetic analysis. The calculated reduction degrees based on this approach are in good agreement with the experimental values.

  14. Determining Pu-239 content by resonance transmission analysis using a filtered reactor beam.

    SciTech Connect

    Klann, R. T.

    1998-11-25

    A novel technique has been developed at Argonne National Laboratory to determine the {sup 239}Pu content in EBR-II blanket elements using resonance transmission analysis (RTA) with a filtered reactor beam. The technique uses cadmium and gadolinium filters along with a {sup 239}Pu fission chamber to isolate the 0.3 eV resonance in {sup 239}Pu. In the energy range from 0.1 to 0.5 eV, the total microscopic cross-section of {sup 239}Pu is significantly larger than the cross-sections of {sup 238}U and {sup 235}U. This large difference in cross-section allows small amounts of {sup 239}Pu to be detected in uranium samples. Tests using a direct beam from a 250 kW TRIGA reactor have been performed with stacks of depleted uranium and {sup 239}Pu foils. Preliminary measurement results are in good agreement with the predicted results up to about two weight percent of {sup 239}Pu in the sample. In addition, measured {sup 239}Pu masses were in agreement with actual sample masses with uncertainties less than 3.8 percent.

  15. Period meter for reactors

    DOEpatents

    Rusch, Gordon K.

    1976-01-06

    An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

  16. NEUTRONIC REACTOR POWER PLANT

    DOEpatents

    Metcalf, H.E.

    1962-12-25

    This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

  17. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

  18. Improved vortex reactor system

    DOEpatents

    Diebold, James P.; Scahill, John W.

    1995-01-01

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  19. Advanced Test Reactor Tour

    SciTech Connect

    Miley, Don

    2011-01-01

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  20. Advanced Test Reactor Tour

    ScienceCinema

    Miley, Don

    2016-07-12

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  1. NEUTRONIC REACTOR SHIELDING

    DOEpatents

    Borst, L.B.

    1961-07-11

    A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

  2. High solids fermentation reactor

    DOEpatents

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  3. High solids fermentation reactor

    DOEpatents

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-03-02

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  4. Efficient Silicon Reactor

    NASA Technical Reports Server (NTRS)

    Bates, H. E.; Hill, D. M.; Jewett, D. N.

    1983-01-01

    High-purity silicon efficiently produced and transferred by continuous two-cycle reactor. New reactor operates in relatively-narrow temperature rate and uses large surfaces area to minimize heat expenditure and processing time in producing silicon by hydrogen reduction of trichlorosilane. Two cycles of reactor consists of silicon production and removal.

  5. Advances in Tandem Mirror fusion power reactors

    SciTech Connect

    Perkins, L.J.; Logan, B.G.

    1986-05-20

    The Tandem Mirror exhibits several distinctive features which make the reactor embodiment of the principle very attractive: Simple low-technology linear central cell; steady-state operation; high-..beta.. operation; no driven current or disruptions; divertorless operation; direction conversion of end-loss power; low-surface heat loads; and advanced fusion fuel capability. In this paper, we examine these features in connection with two tandem mirror reactor designs, MARS and MINIMARS, and several advanced reactor concepts including the wall-stabilized reactor and the field-reversed mirror. With a novel compact end plug scheme employing octopole stabilization, MINIMARS is expressly designed for short construction times, factory-built modules, and a small (600 MWe) but economic reactor size. We have also configured the design for low radioactive afterheat and inherent/passive safety under LOCA/LOFA conditions, thereby obviating the need for expensive engineered safety systems. In contrast to the complex and expensive double-quadrupole end-cell of the MARS reactor, the compact octopole end-cell of MINIMARS enables ignition to be achieved with much shorter central cell lengths and considerably improves the economy of scale for small (approx.250 to 600 MWe) tandem mirror reactors. Finally, we examine the prospects for realizing the ultimate potential of the tandem mirror with regard to both innovative configurations and novel neutron energy conversion schemes, and stress that advanced fuel applications could exploit its unique reactor features.

  6. Mechanical cutting of irradiated reactor internal components

    SciTech Connect

    Anderson, Michael G.

    2008-01-15

    Mechanical cutting methods to volume reduce and package reactor internal components are now a viable solution for stakeholders challenged with the retirement of first generation nuclear facilities. The recent completion of the removal of the Reactor Vessel Internals (RVI) from within the Sacramento Municipal Utility District's (SMUD) Rancho Seco Nuclear Power Plant demonstrates that unlike previous methods, inclusive of plasma arc and abrasive water-jet cutting, mechanical cutting minimizes exposure to workers, costly water cleanup, and excessive secondary waste generation. Reactor internal components were segmented, packaged, and removed from the reactor building for shipment or storage, allowing the reactor cavity to be drained and follow-on reactor segmentation activities to proceed in the dry state. Area exposure rates at the work positions during the segmentation process were generally 1 mR per hr. Radiological exposure documented for the underwater segmentation processes totaled 13 person rem. The reactor internals weighing 343,000 pounds were segmented into over 200 pieces for maximum shipping package efficiency and produced 5,600 lb of stainless steel chips and shavings which were packaged in void spaces of existing disposal containers, therefore creating no additional disposal volume. Because no secondary waste was driven into suspension in the reactor cavity water, the water was free released after one pass through a charcoal bed and ion exchange filter system. Mechanical cutting techniques are capable of underwater segmentation of highly radioactive components on a large scale. This method minimized radiological exposure and costly water cleanup while creating no secondary waste.

  7. A novel plant protection strategy for transient reactors

    NASA Astrophysics Data System (ADS)

    Bhattacharyya, Samit K.; Lipinski, Walter C.; Hanan, Nelson A.

    The present plant protection system (PPS) has been defined for use in the TREAT-upgrade (TU) reactor for controlled transient operation of reactor-fuel behavior testing under simulated reactor-accident conditions. A PPS with energy-dependent trip set points lowered worst-case clad temperatures by as much as 180 K, relative to the use of conventional fixed-level trip set points. The multilayered multilevel protection strategy represents the state-of-the-art in terrestrial transient reactor protection systems, and should be applicable to multi-MW space reactors.

  8. Validation of NESTLE against static reactor benchmark problems

    SciTech Connect

    Mosteller, R.D.

    1996-02-01

    The NESTLE advanced modal code was developed at North Carolina State University with support from Los Alamos National Laboratory and Idaho National Engineering Laboratory. It recently has been benchmarked successfully against measured data from pressurized water reactors (PWRs). However, NESTLE`s geometric capabilities are very flexible, and it can be applied to a variety of other types of reactors. This study presents comparisons of NESTLE results with those from other codes for static benchmark problems for PWRs, boiling water reactors (BWRs), high-temperature gas-cooled reactors (HTGRs) and CANDU heavy- water reactors (HWRs).

  9. Reactor vessel support system

    DOEpatents

    Golden, Martin P.; Holley, John C.

    1982-01-01

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  10. Nuclear reactor overflow line

    DOEpatents

    Severson, Wayne J.

    1976-01-01

    The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

  11. Thorium fueled reactor

    NASA Astrophysics Data System (ADS)

    Sipaun, S.

    2017-01-01

    Current development in thorium fueled reactors shows that they can be designed to operate in the fast or thermal spectrum. The thorium/uranium fuel cycle converts fertile thorium-232 into fissile uranium-233, which fissions and releases energy. This paper analyses the characteristics of thorium fueled reactors and discusses the thermal reactor option. It is found that thorium fuel can be utilized in molten salt reactors through many configurations and designs. A balanced assessment on the feasibility of adopting one reactor technology versus another could lead to optimized benefits of having thorium resource.

  12. High temperature reactors

    NASA Astrophysics Data System (ADS)

    Dulera, I. V.; Sinha, R. K.

    2008-12-01

    With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements.

  13. Spinning fluids reactor

    DOEpatents

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  14. Reactor water cleanup system

    DOEpatents

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  15. Reactor water cleanup system

    DOEpatents

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  16. Development of a reactor engineering workstation at Seabrook station

    SciTech Connect

    Tremblay, M.A.; Gorski, J.P. ); Gurney, P.V. )

    1992-01-01

    The reactor engineers at Seabrook station are responsible for supporting plant operation with respect to the current reactor core design. Advanced assembly designs, complex reactor core loading patterns, and emphasis on efficient and safe operation puts a greater demand on the reactor engineer. The traditional use of static data constants and coarse core modeling, in light of the more complex fuel and core designs of today, results in less than optimum monitoring and predicting tools for the reactor engineer. The incorporation of an advanced three-dimensional nodal code with thermal feedbacks and detailed spatial modeling along with the ability to follow current operational history on a state-of-the-art workstation provides the reactor engineer with a dynamic core monitoring and predictive tool. This approach allows for more accurate and efficient completion of the reactor engineer's tasks. Yankee Atomic Electric Company (YAEC) is currently in the process of providing advanced reactor physics nodal methods to the reactor engineers at Seabrook station. The scope of this project is to supply a reactor engineering workstation with a simplified user interface to an advanced nodal core model as part of an on-line core monitor/predictor for standard reactor engineering calculations. It uses the Studsvik Core Management System (CMS), which primarily consists of the CASMO-3 cross-section generating code and the SIMULATE-3 three-dimensional two-group nodal reactor analysis code.

  17. MANHATTAN PROJECT B REACTOR HANFORD WASHINGTON [HANFORD'S HISTORIC B REACTOR (12-PAGE BOOKLET)

    SciTech Connect

    GERBER MS

    2009-04-28

    The Hanford Site began as part of the United States Manhattan Project to research, test and build atomic weapons during World War II. The original 670-square mile Hanford Site, then known as the Hanford Engineer Works, was the last of three top-secret sites constructed in order to produce enriched uranium and plutonium for the world's first nuclear weapons. B Reactor, located about 45 miles northwest of Richland, Washington, is the world's first full-scale nuclear reactor. Not only was B Reactor a first-of-a-kind engineering structure, it was built and fully functional in just 11 months. Eventually, the shoreline of the Columbia River in southeastern Washington State held nine nuclear reactors at the height of Hanford's nuclear defense production during the Cold War era. The B Reactor was shut down in 1968. During the 1980's, the U.S. Department of Energy began removing B Reactor's support facilities. The reactor building, the river pumphouse and the reactor stack are the only facilities that remain. Today, the U.S. Department of Energy (DOE) Richland Operations Office offers escorted public access to B Reactor along a designated tour route. The National Park Service (NPS) is studying preservation and interpretation options for sites associated with the Manhattan Project. A draft is expected in summer 2009. A final report will recommend whether the B Reactor, along with other Manhattan Project facilities, should be preserved, and if so, what roles the DOE, the NPS and community partners will play in preservation and public education. In August 2008, the DOE announced plans to open B Reactor for additional public tours. Potential hazards still exist within the building. However, the approved tour route is safe for visitors and workers. DOE may open additional areas once it can assure public safety by mitigating hazards.

  18. High temperature catalytic membrane reactors

    SciTech Connect

    Not Available

    1990-03-01

    Current state-of-the-art inorganic oxide membranes offer the potential of being modified to yield catalytic properties. The resulting modules may be configured to simultaneously induce catalytic reactions with product concentration and separation in a single processing step. Processes utilizing such catalytically active membrane reactors have the potential for dramatically increasing yield reactions which are currently limited by either thermodynamic equilibria, product inhibition, or kinetic selectivity. Examples of commercial interest include hydrogenation, dehydrogenation, partial and selective oxidation, hydrations, hydrocarbon cracking, olefin metathesis, hydroformylation, and olefin polymerization. A large portion of the most significant reactions fall into the category of high temperature, gas phase chemical and petrochemical processes. Microporous oxide membranes are well suited for these applications. A program is proposed to investigate selected model reactions of commercial interest (i.e. dehydrogenation of ethylbenzene to styrene and dehydrogenation of butane to butadiene) using a high temperature catalytic membrane reactor. Membranes will be developed, reaction dynamics characterized, and production processes developed, culminating in laboratory-scale demonstration of technical and economic feasibility. As a result, the anticipated increased yield per reactor pass economic incentives are envisioned. First, a large decrease in the temperature required to obtain high yield should be possible because of the reduced driving force requirement. Significantly higher conversion per pass implies a reduced recycle ratio, as well as reduced reactor size. Both factors result in reduced capital costs, as well as savings in cost of reactants and energy.

  19. Neutron flux spectra and radiation damage parameters for the Russian Bor-60 and SM-2 reactors

    SciTech Connect

    Karasiov, A.V.; Greenwood, L.R.

    1995-04-01

    The objective is to compare neutron irradiation conditions in Russian reactors and similar US facilities. Neutron fluence and spectral information and calculated radiation damage parameters are presented for the BOR-60 (Fast Experimental Reactor - 60 MW) and SM-2 reactors in Russia. Their neutron exposure characteristics are comparable with those of the Experimental Breeder Reactor (ERB-II), the Fast Flux Test Facility (FFTF), and the High Flux Isotope Reactor (HFIR) in the United States.

  20. Advanced Demonstration and Test Reactor Options Study

    SciTech Connect

    Petti, David Andrew; Hill, R.; Gehin, J.; Gougar, Hans David; Strydom, Gerhard; Heidet, F.; Kinsey, J.; Grandy, Christopher; Qualls, A.; Brown, Nicholas; Powers, J.; Hoffman, E.; Croson, D.

    2017-01-01

    Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy

  1. Neutron fluxes in test reactors

    SciTech Connect

    Youinou, Gilles Jean-Michel

    2017-01-01

    Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

  2. Nuclear reactors built, being built, or planned, 1991

    SciTech Connect

    Simpson, B.

    1992-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  3. HORIZONTAL BOILING REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1958-11-18

    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  4. High energy reactor neutrinos

    NASA Astrophysics Data System (ADS)

    Raper, Neill

    We present the first measurement of a nonzero reactor neutrino flux with energies above 8 MeV. Measurements are taken with the Daya Bay Reactor Neutrino Experiments detectors, using the Guangdong Nuclear Power Station as a source. Disagreement between data and theory regarding rate and shape of reactor neutrino spectra have made the need for direct measurement clear. Data are especially useful at high energies, where far fewer isotopes contribute. Neutrino candidates are correlated to reactor power and reactor power is extrapolated to zero in order to separate neutrino events from background. We find evidence of reactor neutrinos up to ˜12.5 MeV at 1.92 sigma above 0 and include a survey of isotopes likely to be contributing neutrinos in this energy range.

  5. Hybrid reactors. [Fuel cycle

    SciTech Connect

    Moir, R.W.

    1980-09-09

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  6. Improved vortex reactor system

    DOEpatents

    Diebold, J.P.; Scahill, J.W.

    1995-05-09

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  7. The Integral Fast Reactor

    SciTech Connect

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab.

  8. NEUTRONIC REACTOR CONTROL

    DOEpatents

    Dreffin, R.S.

    1959-12-15

    A control means for a nuclear reactor is described. Particularly a device extending into the active portion of the reactor consisting of two hollow elements coaxially disposed and forming a channel therebetween, the cross sectional area of the channel increasing from each extremity of the device towards the center thereof. An element of neutron absorbing material is slidably positionable within the inner hollow element and a fluid reactor poison is introduced into the channel defined by the two hollow elements.

  9. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Goett, J.J.

    1961-01-24

    A system is described which includes a neutronic reactor containing a dispersion of fissionable material in a liquid moderator as fuel and a conveyor to which a portion of the dispersion may be passed and wherein the self heat of the slurry evaporates the moderator. Means are provided for condensing the liquid moderator and returning it to the reactor and for conveying the dried fissionable material away from the reactor.

  10. FLOW SYSTEM FOR REACTOR

    DOEpatents

    Zinn, W.H.

    1963-06-11

    A reactor is designed with means for terminating the reaction when returning coolant is below a predetermined temperature. Coolant flowing from the reactor passes through a heat exchanger to a lower reservoir, and then circulates between the lower reservoir and an upper reservoir before being returned to the reactor. Means responsive to the temperature of the coolant in the return conduit terminate the chain reaction when the temperature reaches a predetermined minimum value. (AEC)

  11. University Reactor Sharing Program

    SciTech Connect

    Dr. W.D. Reece

    1999-09-01

    The University Reactor Sharing Program provides funding for reactor experimentation to institutions that do not normally have access to a research reactor. Research projects supported by the program include items such as dating geological material to producing high current super conducting magnets. The funding also gives small colleges and universities the opportunity to use the facility for teaching courses in nuclear processes; specifically neutron activation analysis and gamma spectroscopy.

  12. 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.

  13. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, Juhani

    1996-01-01

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

  14. Remote Reactor Monitoring

    SciTech Connect

    Bernstein, Adam; Dazeley, Steve; Dobie, Doug; Marleau, Peter; Brennan, Jim; Gerling, Mark; Sumner, Matthew; Sweany, Melinda

    2014-10-21

    The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

  15. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, J.

    1996-03-19

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

  16. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOEpatents

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  17. Membrane reactors at Degussa.

    PubMed

    Wöltinger, Jens; Karau, Andreas; Leuchtenberger, Wolfgang; Drauz, Karlheinz

    2005-01-01

    The review covers the development of membrane reactor technologies at Degussa for the synthesis of fine chemicals. The operation of fed-batch or continuous biocatalytic processes in the enzyme membrane reactor (EMR) is well established at Degussa. Degussa has experience of running EMRs from laboratory gram scale up to a production scale of several hundreds of tons per year. The transfer of the enzyme membrane reactor from biocatalysis to chemical catalysis in the chemzyme membrane reactor (CMR) is discussed. Various homogeneous catalysts have been investigated in the CMR, and the scope and limitation of this new technique is discussed.

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

    PubMed

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

    2016-01-01

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

  19. Prospects for Tokamak Fusion Reactors

    SciTech Connect

    Sheffield, J.; Galambos, J.

    1995-04-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.

  20. Russian RBMK reactor design information

    SciTech Connect

    Not Available

    1993-11-01

    This document concerns the systems, design, and operations of the graphite-moderated, boiling, water-cooled, channel-type (RBMK) reactors located in the former Soviet Union (FSU). The Russian Academy of Sciences Nuclear Safety Institute (NSI) in Moscow, Russia, researched specific technical questions that were formulated by the Pacific Northwest Laboratory (PNL) and provided detailed technical answers to those questions. The Russian response was prepared in English by NSI in a question-and-answer format. This report presents the results of that technical exchange in the context they were received from the NSI organization. Pacific Northwest Laboratory is generating this document to support the US Department of Energy (DOE) community in responding to requests from FSU states, which are seeking Western technological and financial assistance to improve the safety systems of the Russian-designed reactors. This report expands upon information that was previously available to the United States through bilateral information exchanges, international nuclear society meetings, International Atomic Energy Agency (IAEA) reactor safety programs, and Research and Development Institute of Power Engineering (RDIPE) reports. The response to the PNL questions have not been edited or reviewed for technical consistency or accuracy by PNL staff or other US organizations, but are provided for use by the DOE community in the form they were received.