WorldWideScience

Sample records for premiere test reactors

  1. Reactor Simulator Testing

    Science.gov (United States)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise J.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  2. Reactor Simulator Testing

    Science.gov (United States)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise Jon

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz. Keywords: fission, space power, nuclear, liquid metal, NaK.

  3. Test reactor risk assessment methodology

    Energy Technology Data Exchange (ETDEWEB)

    Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.

    1976-04-01

    A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor.

  4. FASTER Test Reactor Preconceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-31

    The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  5. PITR: Princeton Ignition Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-01

    The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection.

  6. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  7. Reactor Simulator Integration and Testing

    Science.gov (United States)

    Schoenfield, M. P.; Webster, K. L.; Pearson, J. B.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator (RxSim) test loop was designed and built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing were to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V because the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This Technical Memorandum summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained, which was lower than the predicted 750 K but 156 K higher than the cold temperature, indicating the design provided some heat regeneration. The annular linear induction pump tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  8. Comparing balance and instep angle of premier and non- premier leg in female soccer athletics

    Directory of Open Access Journals (Sweden)

    Mahsa Mohammadzadeh

    2016-03-01

    Full Text Available This study is aimed to compare balanceand instep angle in premier and non- premier leg. The number of 30 female futsal athletics of Tehran Province (in the age of years old, in the weight of kg participated in this study. To determine premier and non- premier leg of subjects, they besides questioning requested to shoot 5 times and the leg with less error recorded as premier leg. Balance Error Scoring System (BESS test was used for evaluating static balance. 4 markers were used for evaluating instep angle according to Clark Model. For evaluating dynamic balance of subjects single-leg landing test on 3- axes force board with 1000 hertz frequency was used and the time to stability was computed in line with anteroposterior and internal- external based on sequential estimation technique. Data analyzing was done in two sections, first, instep angle was compared in different status and static and dynamic balance in premier and non- premier leg by paired t- test and then the relationship between instep angle was examined in different status with static balance (total score of balance in the test (BESS with instep angle in standing on two leg and single leg for premier leg and dynamic balance in 3 anteroposterior (AP, mediolateral (ML and vertical (V sides by using from correlation coefficient test (for normal data of Pearson correlation coefficient and Spearman abnormal data. There is significant difference between dynamic balance in premier and dynamic balance in non- premier leg of futsal athletics and the premier leg has further balance than non- premier leg. There is negative and significant relationship between dynamic balance and instep angle in all statusexcept premier leg of single- leg in vertical side. Futsal athletics in premier leg have further balance than non- premier leg.

  9. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  10. Testing commercial catalysts in recycle reactors

    Energy Technology Data Exchange (ETDEWEB)

    Berty, J.M.

    1979-01-01

    Recycle reactors for quality control of catalyst production and for testing new catalysts for known or new processes have the following advantages over tubular reactors: they can reproduce the physical and chemical regime which surrounds the catalyst in a commercial reactor; they can achieve high mass and heat transfer; they exhibit uniform coke deposit; and they provide independence of mass velocity and space velocity. Their disadvantage is the unconventional specification of experiments in terms of discharge concentration which derives from the implicit nature of the basic mathematical relationships. Recycle reactor test methods are outlined for quality control and for testing catalysts, e.g., supported nickel from different manufacturers, for processes whose chemistry is well known. Approaches for testing catalysts for new processes are discussed. The standard recycle reactor developed at Union Carbide Corp. and manufactured by Autoclave Engineers, and several of its modifications are described.

  11. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  12. Dynamic model of Fast Breeder Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vaidyanathan, G., E-mail: vaidya@igcar.gov.i [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Kasinathan, N.; Velusamy, K. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)

    2010-04-15

    Fast Breeder Test Reactor (FBTR) is a 40 M Wt/13.2 MWe sodium cooled reactor operating since 1985. It is a loop type reactor. As part of the safety analysis the response of the plant to various transients is needed. In this connection a computer code named DYNAM was developed to model the reactor core, the intermediate heat exchanger, steam generator, piping, etc. This paper deals with the mathematical model of the various components of FBTR, the numerical techniques to solve the model, and comparison of the predictions of the code with plant measurements. Also presented is the benign response of the plant to a station blackout condition, which brings out the role of the various reactivity feedback mechanisms combined with a gradual coast down of reactor sodium flow.

  13. Premier Hospital Historical Data

    Data.gov (United States)

    U.S. Department of Health & Human Services — To provide a historical overview of the participating hospitals, before the first project report, Premier Healthcare Informatics has used data already available for...

  14. Advanced Demonstration and Test Reactor Options Study

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Gehin, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States); Qualls, A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Croson, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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

  15. Advanced Test Reactor National Scientific User Facility Progress

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Todd R. Allen; James I. Cole; Jeff B. Benson; Mary Catherine Thelen

    2012-10-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. Goals of the ATR NSUF are to define the cutting edge of nuclear technology research in high temperature and radiation environments, contribute to improved industry performance of current and future light water reactors, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives

  16. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

  17. PREMIER MEETS THE PRESS

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    On March 14, Premier Wen Jiabao addressed the Chinese and foreign media at a press conference after the closing meeting of the Third Session of the 11th National People’s Congress. Edited highlights on a number of economic and social issues follow:

  18. PREMIER MEETS THE PRESS

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    @@ March 14.Premier Wen Jiabao addressed the Chinese and foreign media at a press conference after the closing meeting of the Third Session of the 1 lth National People's Congress.Edited highlights on a number of economic and social issues follow:

  19. Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, C.; Wachs, D.; Carmack, J.; Woolstenhulme, N.

    2017-01-01

    The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, and salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.

  20. Irradiation rigs in material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rozenblum, F.; Gonnier, C.; Bignan, G. [CEA, Research Centers of Saclay and Cadarache (France)

    2011-07-01

    Osiris is a research reactor with a thermal power of 70 MW. It is a light-water reactor, open-core pool type, the principal aim of which is to carry out tests and irradiate structural materials and fuel elements of nuclear power plants under a high flux of neutrons, and to produce radioisotopes. Osiris operates around 200 days a year, in cycles of varying lengths from 3 to 4 weeks. A shutdown of about 10 days between two cycles allows reloading the core with fuel. Mainly 2 types of irradiation device are present: capsules for materials irradiation (CHOUCA and IRMA devices) and fuels irradiation loops (GRIFFONOS and ISABELLE). Although Osiris is still providing experiments of very good quality, it is facing obsolescence due to its ageing. Osiris is planned to be shut down during next decade. Consequently, it has been decided to launch the construction of the Jules Horowitz Reactor (JHR) in Cadarache. JHR is a water cooled reactor which provides the necessary flexibility and accessibility to manage several highly instrumented experiments, reproducing different reactor environments (water, gas or liquid metal loops), generating transient regimes (key for safety). The JHR facility includes the reactor building, including core, cooling system and the experimental bunkers connected to the core through pool wall penetrations and the auxiliary building, including pools and hot cells necessary for the experimental irradiation process. JHR core is optimised to produce high fast neutron flux to study structural material ageing and high thermal neutrons flux for fuel experiments. The conception of this first fleet of devices integrates the operational experience accumulated by the existing MTR and specifically the Osiris one

  1. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG

  2. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  3. Instrumentation to Enhance Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  4. Reevaluation of the Premier Clostridium difficile toxin A and B immunoassay with comparison to glutamate dehydrogenase common antigen testing evaluating Bartels cytotoxin and Prodesse ProGastro Cd polymerase chain reaction as confirmatory procedures.

    Science.gov (United States)

    Doing, Kirk M; Hintz, Marilyn S; Keefe, Calvin; Horne, Sarah; LeVasseur, Shelby; Kulikowski, Martha L

    2010-02-01

    Enzyme immunoassays are currently the most common tests used in the clinical laboratory for the detection of Clostridium difficile toxins; however, significant problems with their performance have recently been described. We prospectively reevaluated the Meridian Premier C. difficile toxin A/B assay with direct comparison to a 2-step algorithm that screened for C. difficile common antigen and compared cytotoxin and real-time polymerase chain reaction (PCR) as confirmatory procedures. The Premier assay lacked sufficient sensitivity, missing 25% of true-positive samples. PCR was the most sensitive method and the only procedure that allowed same day testing and reporting.

  5. C Reactor overbore test facility review

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, P.A.; Nilson, R.

    1964-04-24

    In 1961, large-size, smooth-bore, Zircaloy process tubes were installed in C-Reactor graphite channels that had been enlarged to 2.275 inches. These tubes were installed to provide a test and demonstration facility for the concept of overboring as a means of securing significant improvement in the production capability of the reactors, After two years of facility operation, it is now appropriate to consider the extent to which original objectives have been achieved, to re-examine the original objectives, and to consider the best future use of this unique facility. This report presents the general results of such a review and re-examination in more detail.

  6. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-02-15

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

  7. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  8. Dielectric Heaters for Testing Spacecraft Nuclear Reactors

    Science.gov (United States)

    Sims, William Herbert; Bitteker, Leo; Godfroy, Thomas

    2006-01-01

    A document proposes the development of radio-frequency-(RF)-driven dielectric heaters for non-nuclear thermal testing of the cores of nuclear-fission reactors for spacecraft. Like the electrical-resistance heaters used heretofore for such testing, the dielectric heaters would be inserted in the reactors in place of nuclear fuel rods. A typical heater according to the proposal would consist of a rod of lossy dielectric material sized and shaped like a fuel rod and containing an electrically conductive rod along its center line. Exploiting the dielectric loss mechanism that is usually considered a nuisance in other applications, an RF signal, typically at a frequency .50 MHz and an amplitude between 2 and 5 kV, would be applied to the central conductor to heat the dielectric material. The main advantage of the proposal is that the wiring needed for the RF dielectric heating would be simpler and easier to fabricate than is the wiring needed for resistance heating. In some applications, it might be possible to eliminate all heater wiring and, instead, beam the RF heating power into the dielectric rods from external antennas.

  9. The Next Generation Nuclear Plant Graphite Creep Experiment Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of

  10. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  11. Grey Rod Test in HANARO Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

    2008-08-15

    Westinghouse/KAERI/KNF agreed to perform an irradiation test in the HANARO reactor to obtain irradiation data on the new grey rods that will be part of an AP1000 system. As a preliminary test, two samples containing pure Ag (Reference) and Ag-In-Cd materials provided by Westinghouse Electric Company (WEC) were inserted in a KNF irradiation capsule of 07M-13N. The specimens were irradiated for 95.19days (4 cycles) in the CT test hole of the HANARO of a 30MW thermal output to have a fast neutron fluence of 1.11x10{sup 21}(n/cm{sup 2}) (E>1.0MeV). This report provides all the test conditions and data obtained during the irradiation test of the grey rods in HANARO requested by Westinghouse. The test was prepared according to the meeting minutes (June 26, 2007) and the on-going subject test was stopped midway by the request of Westinghouse.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  13. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

  14. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

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

  15. Preliminary Options Assessment of Versatile Irradiation Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sen, Ramazan Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    The objective of this report is to summarize the work undertaken at INL from April 2016 to January 2017 and aimed at analyzing some options for designing and building a versatile test reactor; the scope of work was agreed upon with DOE-NE. Section 2 presents some results related to KNK II and PRISM Mod A. Section 3 presents some alternatives to the VCTR presented in [ ] as well as a neutronic parametric study to assess the minimum power requirement needed for a 235U metal fueled fast test reactor capable to generate a fast (>100 keV) flux of 4.0 x 1015 n /cm2-s at the test location. Section 4 presents some results regarding a fundamental characteristic of test reactors, namely displacement per atom (dpa) in test samples. Section 5 presents the INL assessment of the ANL fast test reactor design FASTER. Section 6 presents a summary.

  16. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. [ed.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG&G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  17. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. (ed.); Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  18. Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States); Palmer, A.J.; Ingram, F.W. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States); Wiffen, F.W. [Dept. of Energy, Germantown, MD (United States). Office of Fusion Energy

    1998-09-01

    The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

  19. EBR-2 (Experimental Breeder Reactor-2), IFR (Integral Fast Reactor) prototype testing programs

    Energy Technology Data Exchange (ETDEWEB)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W. (Argonne National Lab., Idaho Falls, ID (USA). EBR-II Div. Argonne National Lab., IL (USA)); Planchon, H.P.; Lambert, J.D.B. (Argonne National Lab., IL (USA))

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs.

  20. Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2009-05-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any

  1. Premier Wen on Domestic Policies

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    The Second Session of the 11th National People’s Congress held a press conference at the Great Hall of the People on March 13. Premier Wen Jiabao answered questions from the Chinese and foreign press. Below are highlights of his answers on domestic policies.

  2. Our Measures Are Effective, Premier

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    @@ In an exclusive interview with Xinhua News Agency in Beijing on December 27, 2009, Chinese Premier Wen Jiabao shared his visions of the current economic situation in China and the government's economic policy orientations for 2010. Here are the highlights of the interview.

  3. The RES Reactor. A test reactor for the French naval propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Pivet, Sylvestre [CEA, Centre de Cadarache, F-13108 Saint Paul lez Durance (France); Minguet, Jean-Luc [AREVA-Technicatome, BP17, 91192 Gif-sur-Yvette (France)

    2006-07-01

    In the Cadarache nuclear research centre the French Atomic Energy Commission (CEA) operates, with the support of TECHNICATOME as nuclear operator, the experimental facilities which are necessary for the French naval propulsion program. Since the sixties these facilities have brought a large contribution to the development and to the technical support for the nuclear propulsion; they have been used also to train the French Navy operators. The last experimental reactor, the RNG, is now at the end of its life cycle after thirty years of a profitable operation. A replacement reactor is needed to sustain any evolution of the naval propulsion reactors as well as to guarantee a safe operation and a high level of availability of the existing onboard reactors. The aim of the RES program is namely to build such a test facility. Its construction program started in 2003. By the year 2009 the RES reactor will take over the mission of the RNG. We present hereafter: - A brief history of the French experimental reactors built in support to the naval propulsion, - The needs of the naval propulsion and the related objectives of the RES program, - The corresponding architecture and main characteristics of the RES facility, - The current status of the RES construction. The contents of the paper is as follows: 1. Introduction; 2. History of the French nuclear propulsion experimental reactors; 3. Needs of the naval propulsion and related objectives of the RES reactor; 4. RES architecture and main characteristics; 4.1. The pool module; 4.2. The reactor module; 4.3. The RES reactor, an innovative concept; 5. Realisation status; 6. Conclusion. To summarize, from the year 2009 the RES will be an efficient facility available for irradiation and qualification programs. Its large experimental capabilities will allow relevant fuel and core irradiations. This will give access to a real progress in the knowledge of fuel and core physics as well as in the related simulation tools. This reactor

  4. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  5. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014

    Energy Technology Data Exchange (ETDEWEB)

    Ogden, Dan [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014 Highlights • Rory Kennedy, Dan Ogden and Brenden Heidrich traveled to Germantown October 6-7, for a review of the Infrastructure Management mission with Shane Johnson, Mike Worley, Bradley Williams and Alison Hahn from NE-4 and Mary McCune from NE-3. Heidrich briefed the group on the project progress from July to October 2014 as well as the planned path forward for FY15. • Jim Cole gave two invited university seminars at Ohio State University and University of Florida, providing an overview of NSUF including available capabilities and the process for accessing facilities through the peer reviewed proposal process. • Jim Cole and Rory Kennedy co-chaired the NuMat meeting with Todd Allen. The meeting, sponsored by Elsevier publishing, was held in Clearwater, Florida, and is considered one of the premier nuclear fuels and materials conferences. Over 340 delegates attended with 160 oral and over 200 posters presented over 4 days. • Thirty-one pre-applications were submitted for NSUF access through the NE-4 Combined Innovative Nuclear Research Funding Opportunity Announcement. • Fourteen proposals were received for the NSUF Rapid Turnaround Experiment Summer 2014 call. Proposal evaluations are underway. • John Jackson and Rory Kennedy attended the Nuclear Fuels Industry Research meeting. Jackson presented an overview of ongoing NSUF industry research.

  6. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

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

  7. Preconceptual design of the new production reactor circulator test facility

    Energy Technology Data Exchange (ETDEWEB)

    Thurston, G.

    1990-06-01

    This report presents the results of a study of a new circulator test facility for the New Production Reactor Modular High-Temperature Gas-Cooled Reactor. The report addresses the preconceptual design of a stand-alone test facility with all the required equipment to test the Main Circulator/shutoff valve and Shutdown Cooling Circulator/shutoff valve. Each type of circulator will be tested in its own full flow, full power helium test loop. Testing will cover the entire operating range of each unit. The loop will include a test vessel, in which the circulator/valve will be mounted, and external piping. The external flow piping will include a throttle valve, flowmeter, and heat exchanger. Subsystems will include helium handling, helium purification, and cooling water. A computer-based data acquisition and control system will be provided. The estimated costs for the design and construction of this facility are included. 2 refs., 15 figs.

  8. Final safeguards analysis, high temperature lattice test reactor. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Hanthorn, H.E.; Brown, W.W.; Clark, R.G.; Heineman, R.E.; Humes, R.M.

    1966-01-01

    The PMACS `reactor-normal` signal signifies that important process variables do not exceed their set points, that various interlocks are properly set, that functional tests of the computer operation are satisfactory, and that the reactor flux level and period derived from two additional, independent, and dissimilar channels are within set limits. This safety circuit combines the features of redundancy, dissimilar components, and frequent testing which are required for best reliability. The experimental equipment auxiliary to the reactor includes two oscillator mechanisms, one to move the test cell or the adjoining cell into and out of position, the other to move small specimens in the test cell or adjoining cells. They have cooling chambers for the removal of specimens from the test cell without the necessity of cooling the reactor. A neutron chopper and time-of-flight spectrometer are provided; the neutron detectors, at the end of a 25-meter flight tube, are in an adjoining small building. Test cores may be assembled on a core dolly have a load capacity of 14,000 lb. Two wire traverse mechanisms are provided for measurements of flux distribution.

  9. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  10. The technology development for surveillance test of reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Sun Phil; Park, Day Young; Choi, Kwen Jai

    1997-12-01

    Benchmark test was performed in accordance with the requirement of US NRC Reg. Guide DG-1053 for Kori unit-1 in order to determine best-estimated fast neutron fluence irradiated into reactor vessel. Since the uncertainty of radiation analysis comes from the calculation error due to neutron cross-section data, reactor core geometrical dimension, core source, mesh density, angular expansion and convergence criteria, evaluation of calculational uncertainty due to analytical method was performed in accordance with the regulatory guide and the proof was performed for entire analysis by comparing the measurement value obtained by neutron dosimetry located in surveillance capsule. Best-estimated neutron fluence in reactor vessel was calculated by bias factor, neutron flux measurement value/calculational value, from reanalysis result from previous 1st through 4th surveillance testing and finally fluence prediction was performed for the end of reactor life and the entire period of plant life extension. Pressurized thermal shock analysis was performed in accordance with 10 CFR 50.61 using the result of neutron fluence analysis in order to predict the life of reactor vessel material and the criteria of safe operation for Kori unit 1 was reestablished. (author). 55 refs., 55 figs.

  11. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  12. Materials Test-2 LOCA Simulation in the NRU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Barner, J. O.; Hesson, G. M.; King, I. L.; Marshall, R. K.; Parchen, L. J.; Pilger, J. P.; Rausch, W. N.; Russcher, G. E.; Webb, B. J.; Wildung, N. J.; Wilson, C. L.; Wismer, M. D.; Mohr, C. L.

    1982-03-01

    A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This third experiment of the program produced fuel cladding temperatures exceeding 1033 K (1400°F) for 155 s and resulted in eight ruptured fuel rods. Experiment data and initial results are presented in the form of photographs and graphical summaries.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  15. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1988-05-01

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

  16. Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-15

    A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

  17. Testing piezoelectric sensors in a nuclear reactor environment

    Science.gov (United States)

    Reinhardt, Brian T.; Suprock, Andy; Tittmann, Bernhard

    2017-02-01

    Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this work piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65×1020 nf/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 × 1020 nf/cm2, Zinc Oxide is capable of transduction up to at least 6.27 × 1020 nf/cm2, and Aluminum Nitride is capable of transduction up to at least 8.65 × 1020 nf/cm2.

  18. Enhanced in-pile instrumentation at the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2011-07-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

  19. Injury risk factors, screening tests and preventative strategies: a systematic review of the evidence that underpins the perceptions and practices of 44 football (soccer) teams from various premier leagues.

    Science.gov (United States)

    McCall, Alan; Carling, Chris; Davison, Michael; Nedelec, Mathieu; Le Gall, Franck; Berthoin, Serge; Dupont, Gregory

    2015-05-01

    To systematically review the scientific level of evidence for the 'Top 3' risk factors, screening tests and preventative exercises identified by a previously published survey of 44 premier league football (soccer) teams. Also, to provide an overall scientific level of evidence and graded recommendation based on the current research literature. A systematic literature search (Pubmed [MEDLINE], SportDiscus, PEDRO and Cochrane databases). The quality of the articles was assessed and a level of evidence (1++ to 4) was assigned. Level 1++ corresponded to the highest level of evidence available and 4, the lowest. A graded recommendation (A: strong, B: moderate, C: weak, D: insufficient evidence to assign a specific recommendation) for use in the practical setting was given. Fourteen studies were analysed. The overall level of evidence for the risk factors previous injury, fatigue and muscle imbalance were 2++, 4 and 'inconclusive', respectively. The graded recommendation for functional movement screen, psychological questionnaire and isokinetic muscle testing were all 'D'. Hamstring eccentric had a weak graded 'C' recommendation, and eccentric exercise for other body parts was 'D'. Balance/proprioception exercise to reduce ankle and knee sprain injury was assigned a graded recommendation 'D'. The majority of perceptions and practices of premier league teams have a low level of evidence and low graded recommendation. This does not imply that these perceptions and practices are not important or not valid, as it may simply be that they are yet to be sufficiently validated or refuted by research. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://group.bmj.com/group/rights-licensing/permissions.

  20. Injury risk factors, screening tests and preventative strategies: a systematic review of the evidence that underpins the perceptions and practices of 44 football (soccer) teams from various premier leagues

    Science.gov (United States)

    McCall, Alan; Carling, Chris; Davison, Michael; Nedelec, Mathieu; Le Gall, Franck; Berthoin, Serge; Dupont, Gregory

    2015-01-01

    Purpose To systematically review the scientific level of evidence for the ‘Top 3’ risk factors, screening tests and preventative exercises identified by a previously published survey of 44 premier league football (soccer) teams. Also, to provide an overall scientific level of evidence and graded recommendation based on the current research literature. Methods A systematic literature search (Pubmed [MEDLINE], SportDiscus, PEDRO and Cochrane databases). The quality of the articles was assessed and a level of evidence (1++ to 4) was assigned. Level 1++ corresponded to the highest level of evidence available and 4, the lowest. A graded recommendation (A: strong, B: moderate, C: weak, D: insufficient evidence to assign a specific recommendation) for use in the practical setting was given. Results Fourteen studies were analysed. The overall level of evidence for the risk factors previous injury, fatigue and muscle imbalance were 2++, 4 and ‘inconclusive’, respectively. The graded recommendation for functional movement screen, psychological questionnaire and isokinetic muscle testing were all ‘D’. Hamstring eccentric had a weak graded ‘C’ recommendation, and eccentric exercise for other body parts was ‘D’. Balance/proprioception exercise to reduce ankle and knee sprain injury was assigned a graded recommendation ‘D’. Conclusions The majority of perceptions and practices of premier league teams have a low level of evidence and low graded recommendation. This does not imply that these perceptions and practices are not important or not valid, as it may simply be that they are yet to be sufficiently validated or refuted by research. PMID:25576530

  1. Precision Cleaning - Path to Premier

    Science.gov (United States)

    Mackler, Scott E.

    2008-01-01

    ITT Space Systems Division s new Precision Cleaning facility provides critical cleaning and packaging of aerospace flight hardware and optical payloads to meet customer performance requirements. The Precision Cleaning Path to Premier Project was a 2007 capital project and is a key element in the approved Premier Resource Management - Integrated Supply Chain Footprint Optimization Project. Formerly precision cleaning was located offsite in a leased building. A new facility equipped with modern precision cleaning equipment including advanced process analytical technology and improved capabilities was designed and built after outsourcing solutions were investigated and found lacking in ability to meet quality specifications and schedule needs. SSD cleans parts that can range in size from a single threaded fastener all the way up to large composite structures. Materials that can be processed include optics, composites, metals and various high performance coatings. We are required to provide verification to our customers that we have met their particulate and molecular cleanliness requirements and we have that analytical capability in this new facility. The new facility footprint is approximately half the size of the former leased operation and provides double the amount of throughput. Process improvements and new cleaning equipment are projected to increase 1st pass yield from 78% to 98% avoiding $300K+/yr in rework costs. Cost avoidance of $350K/yr will result from elimination of rent, IT services, transportation, and decreased utility costs. Savings due to reduced staff expected to net $4-500K/yr.

  2. LOCA simulation in the NRU reactor: materials test-1

    Energy Technology Data Exchange (ETDEWEB)

    Russcher, G.E.; Marshall, R.K.; Hesson, G.M.; Wildung, N.J.; Rausch, W.N.

    1981-10-01

    A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This second experiment of the program produced peak fuel cladding temperatures of 1148K (1607/sup 0/F) and resulted in six ruptured fuel rods. Test data and initial results from the experiment are presented here in the form of photographs and graphical summaries. These results are also compared with the preceding prototypic thermal-hydraulic test results and with computer model test predictions.

  3. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  4. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  5. The Jules Horowitz reactor (JHR), a European material testing reactor (MTR), with extended experimental capabilities

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation... To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.10{sup 14} ncm{sup -2} s{sup -1} and a fast flux of 6,4.10{sup 14} ncm{sup -2}s{sup -1}, it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = displacement per atom.) The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (authors)

  6. The Jules Horowitz reactor (JHR), a European material testing reactor (MTR), with extended experimental capabilities

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation... To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.10{sup 14} ncm{sup -2} s{sup -1} and a fast flux of 6,4.10{sup 14} ncm{sup -2}s{sup -1}, it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = displacement per atom.) The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (authors)

  7. Irradiation tests of texture controlled cladding for pressurized water reactor in foreign reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ukawa, Kazunori [Hokkaido Electric Power Co., Inc., Sapporo (Japan); Abeta, Sadaaki

    1996-12-01

    PWR electric power companies and makers are promoting a high burn-up program for nuclear fuel aiming at reducing spent fuel. PCI tolerance performance of fuel clad is desirable to promote a high burnup program. It is clear that when we change the direction of a Zircaloy crystal structure by improving a manufacture process (texture control), PCI tolerance performance greatly improved. In this study, an improved clad was burnt in a R2 reactor to 61.5 GWd/t and a power ramp test was carried out in Sweden. Based on irradiation data, power ramp test data and post irradiation examination data, improvement of PCI tolerance performance was confirmed. (author)

  8. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs). DATES...

  9. New results from pulse tests in the CABRI reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, F.; Papin, J.; Haessler, M. [Institut de Proterction et de Surete Nucleaire, Saint Paul Lez Durance (France)] [and others

    1996-03-01

    At the 21st and 22nd WRSM (1,2), the motivation and objectives of the French program on the behaviour of high burnup PWR fuel under RIA conditions in the CABRI test reactor has been presented. The major results of the three first tests of the test matrix were presented and in particular REP-Na1, which failed at an unexpected low level of fuel enthalpy, was exposed to the community of nuclear safety research. At this time, no final understanding was reached for the origin of the failure. This objective is reached now. Two further tests, REP-Na4 and 5, have been performed in 1995, they demonstrated a satisfactory and safe behaviour by resisting to the early phase of severe loading during the RIA pulse test. Further examination work and analytical testing is in progress and the next tests with MOX fuel are being prepared.

  10. Final Physics Report for the Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfe, I. B. [Savannah River Site (SRS), Aiken, SC (United States)

    1956-06-25

    This report is a summary of the physics design work performed on the Engineering Test Reactor. The ETR presents computational difficulties not found in other reactors because of the large number of experimental holes in the core. The physics of the ETR depends strongly upon the contents of the in-core experimental facilities. In order to properly evaluate the reactor; taking into account the experiments in the core, multi-region, two-dimensional calculations are required. These calculations require .the use of a large computer such as the Remington Rand Univac and are complex and expensive enough to warrant a five-stage program: 1. In the early stages of design, only preliminary two-dimensional calculations were performed .in order to obtain a rough idea of the general behavior of the reactor and its critical mass with tentative experiments in place. 2. A large amount of work was carried out in which the reactor was approximated as one with a uniform homogeneous core. With this model, detailed studies were carried out to investigate the feasibility and to obtain general design data on such points as the design and properties of the gray and black-control rods, the design of the beryllium reflector, gamma and neutron heating, the use of burnable poisons, etc. In performing these calculations, use was made of the IBM 650 PROD code obtained from KAPL. 3. With stages 1 and 2 carried out, two-dimensional calculations of the core at start-up conditions were performed on the Univac computer. 4. Detailed two-dimensional calculations of the properties of the ETR with a proposed first set of experiments in place were carried out. 5. A series of nuclear tests were performed at the reactivity measurements facility at the MTR site in order to confirm the validity of the analytical techniques in physics analysis. In performing the two-dimensional Univac calculations, the MUG code developed by KAPL and the Cuthill code developed at the David Taylor Model Basin were utilized. In

  11. Conceptual design study of a scyllac fusion test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomassen, K.I. (comp.)

    1975-07-01

    The report describes a conceptual design study of a fusion test reactor based on the Scyllac toroidal theta-pinch approach to fusion. It is not the first attempt to describe the physics and technology required for demonstrating scientific feasibility of the approach, but it is the most complete design in the sense that the physics necessary to achieve the device goals is extrapolated from experimentally tested MHD theories of toroidal systems,and it uses technological systems whose engineering performance has been carefully calculated to ensure that they meet the machine requirements.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

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

  13. Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

    Energy Technology Data Exchange (ETDEWEB)

    S. Bragg-Sitton; J. Bess; J. Werner; G. Harms; S. Bailey

    2011-09-01

    Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al., 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).

  14. Local stability tests in Dresden 2 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    March-Leuba, J.; Fry, D.N.; Buchanan, M.E.; McNew, C.O.

    1984-04-01

    This report presents the results of a local stability test performed at Dresden Unit 2 in May 1983 to determine the effect of a new fuel element design on local channel stability. This test was performed because the diameter of the new fuel rods increases the heat transfer coefficient, making the reactor more responsive and, thus, more susceptible to instabilities. After four of the new fuel elements with a 9 x 9 array of fuel rods were loaded into Dresden 2, the test was performed by inserting an adjacent control rod all the way in and then withdrawing it to its original position at maximum speed. At the moment of the test, reactor conditions were 52.7% power and 38.9% flow. Both the new 9 x 9 fuel elements and the standard 8 x 8 ones proved to be locally stable when operating at minimum pump speed at the beginning of cycle in Dresden 2, and no significant difference was found between the behavior of the two fuel types. Finally, Dresden 2 showed a high degree of stability during control rod and normal noise type perturbations.

  15. Safety Analyses at the Idaho National Engineering and Environmental Laboratory Test Reactor Area - Past to Present

    Energy Technology Data Exchange (ETDEWEB)

    Ambrosek, Richard Garry; Ingram, Frederick William

    1999-11-01

    Test reactors are unique in that the core configuration may change with each operating interval. The process of safety analyses for test reactors at the Idaho National Engineering and Environmental Test Reactor Area has evolved as the computing capabilities, software, and regulatory requirements have changed. The evaluations for experiments and the reactor have moved from measurements in a set configuration and then application to other configurations with a relatively large error to modeling in three-dimensions and explicit analyses for each experiment and operating interval. This evolution is briefly discussed for the Test Reactor Area.

  16. Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Culp

    2007-01-26

    This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.

  17. High temperature indentation tests on fusion reactor candidate materials

    Energy Technology Data Exchange (ETDEWEB)

    Montanari, R. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy)]. E-mail: roberto.montanari@uniroma2.it; Filacchioni, G. [ENEA CR Casaccia, Via Anguillarese 301, I-00060 S.M. di Galeria, Rome (Italy); Iacovone, B. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy); Plini, P. [Dipartimento di Ingegneria Meccanica, Universita di Roma-Tor Vergata, Via del Politecnico 1, I-00133 Rome (Italy); Riccardi, B. [Associazione EURATOM-ENEA sulla Fusione, P.O. Box 65, I-00044 Frascati, Rome (Italy)

    2007-08-01

    Flat-top cylinder indenter for mechanical characterization (FIMEC) is an indentation technique employing cylindrical punches with diameters ranging from 0.5 to 2 mm. The test gives pressure-penetration curves from which the yield stress can be determined. The FIMEC apparatus was developed to test materials in the temperature range from -180 to +200 {sup o}C. Recently, the heating system of FIMEC apparatus has been modified to operate up to 500 {sup o}C. So, in addition to providing yield stress over a more extended temperature range, it is possible to perform stress-relaxation tests at temperatures of great interest for several nuclear fusion reactor (NFR) alloys. Data on MANET-II, F82H mod., Eurofer-97, EM-10, AISI 316 L, Ti6Al4V and CuCrZr are presented and compared with those obtained by mechanical tests with standard methods.

  18. Data on test results of vessel cooling system of high temperature engineering test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Saikusa, Akio [Secretariat of Nuclear Safety Commission, Tokyo (Japan); Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-02-01

    High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

  19. Entrained Flow Reactor Test of Potassium Capture by Kaolin

    DEFF Research Database (Denmark)

    Wang, Guoliang; Jensen, Peter Arendt; Wu, Hao

    2015-01-01

    In the present study a method to simulate the reaction between gaseous KCl and kaolin at suspension fired condition was developed using a pilot-scale entrained flow reactor (EFR). Kaolin was injected into the EFR for primary test of this method. By adding kaolin, KCl can effectively be captured......, forming water-insoluble K-aluminosilicate. The amount of K captured by 1 g kaolin rose when increasing the molar ratio of K/Si in the reactant. Changing of reaction temperature from 1100 °C to 1300 °C did not influence the extent of reaction, which is different from the results observed in previous fixed...

  20. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  1. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

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

  2. EPR/PTFE dosimetry for test reactor environments

    Energy Technology Data Exchange (ETDEWEB)

    Vehar, D.W.; Griffin, P.J.; Quirk, T.J. [Sandia National Laboratories, Albuquerque, NM 87185-1146 (United States)

    2011-07-01

    The use of Electron Paramagnetic Resonance (EPR) spectroscopy with materials such as alanine is well established as a technique for measurement of ionizing radiation absorbed dose in photon and electron fields such as Co-60, high-energy bremsstrahlung and electron-beam fields [1]. In fact, EPR/Alanine dosimetry has become a routine transfer standard for national standards bodies such as NIST and NPL. In 1992 the Radiation Metrology Laboratory (RML) at Sandia National Laboratories implemented EPR/Alanine capabilities for use in routine and calibration activities at its Co-60 and pulsed-power facilities. At that time it also investigated the usefulness of the system for measurement of absorbed dose in the mixed neutron/photon environments of reactors such as the Sandia Pulsed Reactor and the Annular Core Research Reactor used for hardness testing of electronics. The RML concluded that the neutron response of alanine was a sufficiently high fraction of the overall dosimeter response that the resulting uncertainties in the photon dose would be unacceptably large for silicon-device testing. However, it also suggested that non-hydrogenous materials such as polytetrafluoroethylene (PTFE) would exhibit smaller neutron response and might be useful in mixed environments. Preliminary research with PTFE in photon environments indicated considerable promise, but further development was not pursued at that time. Because of renewed interest in absorbed dose measurements that could better define the individual contributions of photon and neutron components to the overall dose delivered to a test object, the RML has re-initiated the development of an EPR/PTFE dosimetry system. This effort consists of three stages: 1) Identification of PTFE materials that may be suitable for dosimetry applications. It was speculated that the inconsistency of EPR signatures in the earlier samples may have been due to variability in PTFE manufacturing processes. 2) Characterization of dosimetry in

  3. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, William Jonathan [Idaho National Laboratory; Barrett, Kristine Eloise [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  4. Meso-scale modeling of irradiated concrete in test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giorla, A. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Vaitová, M. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic); Le Pape, Y., E-mail: lepapeym@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Štemberk, P. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic)

    2015-12-15

    Highlights: • A meso-scale finite element model for irradiated concrete is developed. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • Confrontation with expansion and damage obtained from experiments is successful. • Effects of paste shrinkage, creep and ductility are discussed. - Abstract: A numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale is detailed in this paper. Irradiation experiments in test reactor (Elleuch et al., 1972), i.e., in accelerated conditions, are simulated. Concrete is considered as a two-phase material made of elastic inclusions (aggregate) subjected to thermal and irradiation-induced swelling and embedded in a cementitious matrix subjected to shrinkage and thermal expansion. The role of the hardened cement paste in the post-peak regime (brittle-ductile transition with decreasing loading rate), and creep effects are investigated. Radiation-induced volumetric expansion (RIVE) of the aggregate cause the development and propagation of damage around the aggregate which further develops in bridging cracks across the hardened cement paste between the individual aggregate particles. The development of damage is aggravated when shrinkage occurs simultaneously with RIVE during the irradiation experiment. The post-irradiation expansion derived from the simulation is well correlated with the experimental data and, the obtained damage levels are fully consistent with previous estimations based on a micromechanical interpretation of the experimental post-irradiation elastic properties (Le Pape et al., 2015). The proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-01

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

  8. VISTA : thermal-hydraulic integral test facility for SMART reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, K. Y.; Park, H. S.; Cho, S.; Park, C. K.; Lee, S. J.; Song, C. H.; Chung, M. K. [KAERI, Taejon (Korea, Republic of)

    2003-07-01

    Preliminary performance tests were carried out using the thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), which has been constructed to simulate the SMART-P. The VISTA facility is an integral test facility including the primary and secondary systems as well as safety-related Passive Residual Heat Removal (PRHR) systems. Its scaled ratio with respect to the SMART-P is 1/1 in height and 1/96 in volume and heater power. Several steady states and power changing tests have been carried out to verify the overall thermal hydraulic primary and secondary characteristics in the range of 10% to 100% power operation. As for the preliminary results, the steady state conditions were found to coincide with the expected design values of the SMART-P. But the major thermal hydraulic parameters are greatly affected by the initial water level and the nitrogen pressure in the reactor's upper annular cavity. The power step/ramp changing tests are successfully carried out and the system responses are observed. The primary natural circulation operation is achieved, but advanced control logics need to be developed to reach the natural circulation mode without pressure excursion. In the PRHR transient tests, the natural circulation flow rate through the PRHR system was found to be about 10 percent in the early phases of PRHR operation.

  9. BENCHMARK EVALUATION OF THE START-UP CORE REACTOR PHYSICS MEASUREMENTS OF THE HIGH TEMPERATURE ENGINEERING TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    John Darrell Bess

    2010-05-01

    The benchmark evaluation of the start-up core reactor physics measurements performed with Japan’s High Temperature Engineering Test Reactor, in support of the Next Generation Nuclear Plant Project and Very High Temperature Reactor Program activities at the Idaho National Laboratory, has been completed. The evaluation was performed using MCNP5 with ENDF/B-VII.0 nuclear data libraries and according to guidelines provided for inclusion in the International Reactor Physics Experiment Evaluation Project Handbook. Results provided include updated evaluation of the initial six critical core configurations (five annular and one fully-loaded). The calculated keff eigenvalues agree within 1s of the benchmark values. Reactor physics measurements that were evaluated include reactivity effects measurements such as excess reactivity during the core loading process and shutdown margins for the fully-loaded core, four isothermal temperature reactivity coefficient measurements for the fully-loaded core, and axial reaction rate measurements in the instrumentation columns of three core configurations. The calculated values agree well with the benchmark experiment measurements. Fully subcritical and warm critical configurations of the fully-loaded core were also assessed. The calculated keff eigenvalues for these two configurations also agree within 1s of the benchmark values. The reactor physics measurement data can be used in the validation and design development of future High Temperature Gas-cooled Reactor systems.

  10. First remarks on the initiative of conception and evaluation of the safety of the reactors of the fourth generation; Premieres reflexions sur la demarche de conception et d'evaluation de la surete des reacteurs de quatrieme generation

    Energy Technology Data Exchange (ETDEWEB)

    Clement, Ch.; Maliverney, B.; Mulet-Marquis, D.; Sauvage, J.F. [Electricite de France (EDF), 75 - Paris (France); Guesdon, B. [Technical Secretary (France); Carluec, B.; Ehster, S. [AREVA NP, 92 - Paris la Defense (France); Greneche, D. [AREVA NC, 78 - Velizy Villacoublay (France); Anzieu, P. [CEA Saclay, 91 - Gif sur Yvette (France); Fiorini, G.L. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Rozenholc, M.; Vitton, F.; Rouyer, J.L

    2007-09-15

    The objectives of the nuclear systems of the fourth generation impose ambitious requirements which concern the sustainable development, the competitiveness, the safety, the reliability, the resistance to proliferation and the physical protection. The examination of the policies of safety applicable to these systems is led at the same moment to the national and international levels. In France, thoughts on the safety is organized within the French consultative group of safety (G.C.F.S.) which collects the industrial actors and of research and development involved in the development of the new systems. Within the framework of the initiatives of international harmonization, the G.C.F.S. suggests defining recommendations common to all the concepts of the fourth generation reactors then on the basis of this neutral frame towards the technologies, establishing specific recommendations for each of the systems. (N.C.)

  11. Development and verification test of integral reactor major components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. I.; Kim, Y. W.; Kim, J. H. and others

    1999-03-01

    The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability.

  12. Natural radioactive materials at the Arco Reactor Test Site

    Energy Technology Data Exchange (ETDEWEB)

    Singlevich, W; Healy, J W; Paas, H J; Carey, Z E

    1951-05-28

    At the request of the Division of Biology and Medicine of the AEC, the Biophysics Section of the Radiological Sciences Department at Hanford undertook the task of conducting a background survey for naturally occurring radioactive materials in the environs of the Arco Reactor Test Site in Central Idaho. This survey was part of an overall study which included meteorological measurements by the the Air Weather Service, Geological Studies by the USGS, and an ecological survey of plants and animals by members of the Idaho State College at Pocatello. In general, the measurements at Arco followed the pattern established for environmental monitoring at the Hanford Site with some additional measurements made for natural isotopes not normally of concern at Hanford. A number of analysis included materials such as plutonium and I-131 which were carried out for the purpose of establishing analytical backgrounds for the procedures used. 20 refs., 13 figs., 11 tabs.

  13. Recent results on the RIA test in IGR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L. [Nuclear Safety Institute, Moscow (Russian Federation)

    1997-01-01

    At the 23d WRSM meeting the data base characterizing results of VVER high burnup fuel rods tests under reactivity-initiated accident (RIA) conditions was presented. Comparison of PWR and VVER failure thresholds was given also. Additional analysis of the obtained results was being carried out during 1996. The results of analysis show that the two different failure mechanisms were observed for PWR and VVER fuel rods. Some factors which can be as the possible reasons of these differences are presented. First of them is the state of preirradiated cladding. Published test data for PWR high burnup fuel rods demonstrated that the PWR high burnup fuel rods failed at the RIA test are characterized by very high level of oxidation and hydriding for the claddings. Corresponding researches were performed at Institute of Atomic Reactors (RLAR, Dimitrovgrad, Russia) for large set of VVER high burnup fuel rods. Results of these investigations show that preirradiated commercial Zr-1%Nb claddings practically keep their initial levels of oxidation and H{sub 2} concentration. Consequently the VVER preirradiated cladding must keep the high level of mechanical properties. The second reason leading to differences between failure mechanisms for two types of high burnup fuel rods can be the test conditions. Now such kind of analysis have been performed by two methods.

  14. Deterministic Modeling of the High Temperature Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, J.; Cogliati, J. J.; Pope, M. A.; Ferrer, R. M.; Ougouag, A. M.

    2010-06-01

    Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL’s current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green’s Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2–3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the

  15. Miniaturized Charpy test for reactor pressure vessel embrittlement characterization

    Energy Technology Data Exchange (ETDEWEB)

    Manahan, M.P. Sr. [MPM Research and Consulting, Lemont, PA (United States)

    1999-10-01

    Modifications were made to a conventional Charpy machine to accommodate the miniaturized Charpy V-Notch (MCVN) specimens which were fabricated from an archived reactor pressure vessel (RPV) steel. Over 100 dynamic MCVN tests were performed and compared to the results from conventional Charpy V-Notch (CVN) tests to demonstrate the efficacy of the miniature specimen test. The optimized sidegrooved MCVN specimens exhibit transitional fracture behavior over essentially the same temperature range as the CVN specimens which indicates that the stress fields in the MCVN specimens reasonably simulate those of the CVN specimens and this fact has been observed in finite element calculations. This result demonstrates a significant breakthrough since it is now possible to measure the ductile-brittle transition temperature (DBTT) using miniature specimens with only small correction factors, and for some materials as in the present study, without the need for any correction factor at all. This development simplifies data interpretation and will facilitate future regulatory acceptance. The non-sidegrooved specimens yield energy-temperature data which is significantly shifted downward in temperature (non-conservative) as a result of the loss of constraint which accompanies size reduction.

  16. Testing of an advanced thermochemical conversion reactor system

    Science.gov (United States)

    1990-01-01

    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions.

  17. Testing of an advanced thermochemical conversion reactor system

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

  18. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  19. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  20. CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Mieleszczenko, W.; Lechniak, J.; Moldysz, A.; Andrzejewski, K.; Kulikowska, T.; Marcinkowska, A.; Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division; Institute of Atomic Energy (Poland)

    2010-03-01

    The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

  1. Design considerations of the irradiation test vehicle for the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    An irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) is being jointly developed by the Lockheed Martin Idaho Technologies Company (LMIT) and the U.S. Fusion Program. The vehicle is intended for neutron irradiation testing of candidate structural materials, including vanadium-based alloys, silicon carbide composites, and low activation steels. It could possibly be used for U.S./Japanese collaboration in the Jupiter Program. The first test train is scheduled to be completed by September 1998. In this report, we present the functional requirements for the vehicle and a preliminary design that satisfies these requirements.

  2. Multi-Purpose Thermal Hydraulic Loop: Advanced Reactor Technology Integral System Test (ARTIST) Facility for Support of Advanced Reactor Technologies

    Energy Technology Data Exchange (ETDEWEB)

    James E. O' Brien; Piyush Sabharwall; SuJong Yoon

    2001-11-01

    Effective and robust high temperature heat transfer systems are fundamental to the successful deployment of advanced reactors for both power generation and non-electric applications. Plant designs often include an intermediate heat transfer loop (IHTL) with heat exchangers at either end to deliver thermal energy to the application while providing isolation of the primary reactor system. In order to address technical feasibility concerns and challenges a new high-temperature multi-fluid, multi-loop test facility “Advanced Reactor Technology Integral System Test facility” (ARTIST) is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water. Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed

  3. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  4. Mastering Adobe Premiere Pro CS6

    CERN Document Server

    Ekert, Paul

    2013-01-01

    Designed to be practical and engaging, Mastering Adobe Premiere Pro CS6 is a project-based book to help you truly augment your skills and become a film editing hotshot.If you're just starting out or even migrating from existing video editing software, then this book is for you. With rapid progression through practical examples constructed to be both engaging and useful, Mastering Adobe Premiere Pro CS6 is ideal for learning the sometimes complex workflows of this powerful application.

  5. Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R., E-mail: rrr@cdtn.br, E-mail: amir@cdtn.br, E-mail: edson@cdtn.br, E-mail: maritzargual@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

  6. Development of automatic Ultrasonic testing equipment for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kor R.; Kim, Jae H.; Lee, Jae C.

    1996-06-01

    The selected weld areas of a reactor pressure vessel and adjacent piping are examined by the remote mechanized ultrasonic testing (MUT) equipment. Since the MUT equipment was purchased from southwest Research Institute (SwRI) in April 1985, 15 inservice inspections and 5 preservice inspections are performed with this MUT equipment. However due to the old age of the equipment and frequent movements to plant sites, the reliability of examination was recently decreased rapidly and it is very difficult to keep spare parts. In order to resolve these problems and to meet the strong request from plant sites, we intend to develop a new 3-axis control system including hardware and software. With this control system, we expect more efficient and reliable examination of the nozzle to shell weld areas, which is specified in ASME Code Section XI. The new 3-axis control system hardware and software were designed and development of our own control system, the advanced technologies of computer control mechanism were established and examination reliability of the nozzle to shell weld area was improved. With the development of our 3-axis control system for PaR ISI-2 computer control system, the reliability of nozzle to shell weld area examination has been improved. The established technologies from the development and detailed analysis of existing control system, are expected to be applied to the similar control systems in nuclear power plants. (author). 12 refs., 4 tabs., 33 figs.

  7. Théorie visco-élastique non-extensive IV. Premiers tests expérimentaux de la théorie simplifiée à modes translationnels

    Science.gov (United States)

    Volino, F.

    This paper describes the first tests of the non-extensive visco-elastic theory with translational modes developed in (III), using viscosity, self-diffusion, shear elasticity, thermal conductivity and surface tension literature data of several liquids, namely water, organic solvents and a nematics, for which numerous data exist in the literature. As in the rotational case (I, II), a remarkable qualitative and quantitative agreement is obtained. The influence of translational-rotational coupling on the flow is examined for the nematics and for water. The obtained results, of equivalent importance as those obtained previouly in (II), are more spectacular in the sense that they concern more common substances. If the proposed description is accepted, important changes in the meaning of some physical concepts are necessarily implied. Cet article décrit les premiers tests de la théorie visco-élastique non-extensive à modes translationnels développée dans l'article (III), à l'aide de données de viscosité, de coefficients d'auto-diffusion, d'élasticité de cisaillement, de conductivité thermique et de tension superficielle, d'un certain nombre de liquides, à savoir l'eau, des solvants organiques usuels et un nématique, pour lesquels de très nombreuses données expérimentales existent dans la littérature. Comme dans le cas rotationnel avec les nématiques (I, II), un accord qualitatif et surtout quantitatif remarquable est obtenu. L'influence du couplage translation-orientation sur les propriétés d'écoulement est examiné dans le cas du nématique, et dans une certaine mesure, de l'eau. Ces résultats, de même importance absolue que ceux obtenus dans (II) avec les aspects rotationnels, sont relativement plus spectaculaires dans la mesure où ils sont relatifs à des substances beaucoup plus communes. Si ce type de description est accepté, ils impliquent inévitablement des changements importants sur la signification de certains concepts physiques.

  8. Our Measures Are Effective,Premier

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    In an exclusive interview with Xinhua News Agency in Beijing on December 27,2009,Chinese Premier Wen Jiabao shared his visions of the current economic situation in China and the government’s economic policy orientations for 2010.Here are the highlights of the interview.

  9. Scottish Premier League Reading Stars Evaluation Report

    Science.gov (United States)

    National Literacy Trust, 2009

    2009-01-01

    Scottish Premier League (SPL) Reading Stars uses the motivational power of football to attract families who need support with literacy into a positive and friendly learning environment. It ran for the first time between March and August 2009 and attracted 225 children and 190 adults to take part in a series of inspirational learning sessions in 23…

  10. Cool Wool Iaunched at Premiere Vision

    Institute of Scientific and Technical Information of China (English)

    2012-01-01

    Visitors to Premiere Vision were greeted by the inspired The Woolmark Company image, rapidly becoming iconic, of a welcoming flock of stylish Merino sheep in sunglasses. It raised many a smile and set the scene for a feel- good exhibition, putting Cool Wool firmly at the doorway to the new summer season 2013.

  11. Pakistanis Eager to Welcome Premier Li Keqiang

    Institute of Scientific and Technical Information of China (English)

    Syed; Ali; Nawaz; Gilani

    2013-01-01

    <正>"We the Pakistani Nation,are very much eager to welcome Chinese Premier Li Keqiang on his first visit to Pakistan after taking over the Premiership of the People’s Republic of China whom we are enjoying deep rooted and exemplary friendship for

  12. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  13. Premier Wen hails sci-tech cooperation with CERN

    CERN Multimedia

    2004-01-01

    Premier Wen Jiabao met CERN's director general Dr Robert Aymar and physicist and Nobel laureate Dr Samuel Chao Chung Ting. Premier Wen emphasied the importance for China to collaborate on fundamental science (0.5 page)

  14. Reactor coolant pump testing using motor current signatures analysis

    Energy Technology Data Exchange (ETDEWEB)

    Burstein, N.; Bellamy, J.

    1996-12-01

    This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

  15. Adaptation of Crack Growth Detection Techniques to US Material Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis; Joy L. Rempe; Gordon Kohse; Yakov Ostrovsky; David M. Carpenter

    2014-04-01

    A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some materials testing reactors (MTRs) outside the U.S., such as the Halden Boiling Water Reactor (HBWR), have deployed a technique to measure crack growth propagation during irradiation. This technique incorporates a compact loading mechanism to stress the specimen during irradiation. A crack in the specimen is monitored using the Direct Current Potential Drop (DCPD) method. A project is underway to develop and demonstrate the performance of a similar type of test rig for use in U.S. MTRs. The first year of this three year project was devoted to designing, analyzing, fabricating, and bench top testing a mechanism capable of applying a controlled stress to specimens while they are irradiated in a pressurized water loop (simulating PWR reactor conditions). During the second year, the mechanism will be tested in autoclaves containing high pressure, high temperature water with representative water chemistries. In addition, necessary documentation and safety reviews for testing in a reactor environment will be completed. In the third year, the assembly will be tested in the Massachusetts Institute of Technology Reactor (MITR) and Post Irradiation Examinations (PIE) will be performed.

  16. In-core materials testing under LWR conditions in the Halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A. [OECD Halden Reactor Project (Norway)

    2002-07-01

    The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

  17. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

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

  18. Le Premier Amendement : un mythe

    Directory of Open Access Journals (Sweden)

    Claude‑Jean Bertrand

    2006-03-01

    Full Text Available Ce papier, plein de verve, d’humeur et d’humour, pose, comme sait si bien le faire l’auteur, des questions profondes sous une apparence paradoxale. Il a été présenté lors d’un colloque sur le Premier amendement organisé à l’Université Lumière-Lyon 2 les 17 et 18 janvier 2003. Certaines communications ont été publiées dans le volume XXIV, n°1 (2003, « Le premier amendement : un modèle américain des libertés » (sous la direction de Vincent Michelot de la Revue Tocqueville.

  19. Study of fast reactor safety test facilities. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods. (DG)

  20. MicroEnterprise Americas: Premiere Issue, 2001

    OpenAIRE

    Inter-American Development Bank (IDB)

    2001-01-01

    This premiere issue of MicroEnterprise Americas concentrates on the microfinance industry, a thriving segment of the Latin American financial sector that has rapidly expanded in the past five years. This issue explores looks at how market leaders have developed technologies, attracted investments, and developed tools for mitigating risk in the difficult financial climate of the past two years. MicroEnterprise Americas is published by the Inter-American Forum on Microenterprise, an annual even...

  1. MicroEnterprise Americas: Premiere Issue, 2001

    OpenAIRE

    Inter-American Development Bank (IDB)

    2001-01-01

    This premiere issue of MicroEnterprise Americas concentrates on the microfinance industry, a thriving segment of the Latin American financial sector that has rapidly expanded in the past five years. This issue explores looks at how market leaders have developed technologies, attracted investments, and developed tools for mitigating risk in the difficult financial climate of the past two years. MicroEnterprise Americas is published by the Inter-American Forum on Microenterprise, an annual even...

  2. IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    I. Glagolenko; D. Wachs; N. Woolstenhulme; G. Chang; B. Rabin; C. Clark; T. Wiencek

    2010-10-01

    Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily due to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.

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

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

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

  4. An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

    2009-11-01

    This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

  5. Fatigue Test of Domestic Manufactured Reactor Pressure Vessel Steel

    Institute of Scientific and Technical Information of China (English)

    ZHONG; Wei-hua; TONG; Zhen-feng; NING; Guang-sheng; YU; Bin-tao

    2013-01-01

    The CAP1400 will be built by our country,after the self-dependent innovation work on the imported technology of AP1000,which is a 3rd generation NPP.Now,the design of CAP1400 key equipment is ongoing,and the fatigue design of the domestic manufactured key equipment,such as reactor pressure vessel(RPV),is found to be a main problem in the design work,as the fatigue data is lacked.Thus the

  6. 10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

    Energy Technology Data Exchange (ETDEWEB)

    Bohachek, Randolph Charles [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-09-01

    The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactors is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.

  7. Mixed oxide fuels testing in the advanced test reactor to support plutonium disposition

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M.; Sterbentz, J.W.; Chang, G.S. [and others

    1995-09-01

    An intense worldwide effort is now under way to find means of reducing the stockpile of weapons-grade plutonium. One of the most attractive solutions would be to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PUO{sub 2}) mixed with urania (UO{sub 2}). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification, (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania, (3) The effects of WGPu isotopic composition, (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight, (5) The effects of americium and gallium in WGPu, (6) Fission gas release from MOX fuel pellets made from WGPu, (7) Fuel/cladding gap closure, (8) The effects of power cycling and off-normal events on fuel integrity, (9) Development of radial distributions of burnup and fission products, (10) Power spiking near the interfaces of MOX and urania fuel assemblies, and (11) Fuel performance code validation. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  9. Moderated heat pipe thermionic reactor (MOHTR) module development and test

    Science.gov (United States)

    Merrigan, Michael A.; Trujillo, Vincent L.

    1992-01-01

    The Moderated Heat Pipe Thermionic Reactor (MOHTR) thermionic space reactor design combines the low risk technology associated with the Thermionic Fuel Element (TFE) Verification Program with the high reliability heat transfer capability of liquid metal heat pipe technology. The resulting design concept, capable of implementation over the power range of 10 to 100 kWe, offers efficiency and reliability with reduced risk of single point failures. The union of TFE and heat pipe technology is achieved by imbedding TFEs and heat pipes in a beryllium matrix to which they are thermally coupled by brazing or by liquid metal (NaK or Na) bonding. The reactor employs an array of TFE modules, each comprising a TFE, a zirconium hydride (ZrH) cylinder for neutron moderation, and heat pipes for transport of heat from the collector surface of the TFE to the waste heat radiator. An advantage of the design is the low temperature drop from the collector surface to the radiating surface. This is a result of the elimination of electrical insulation from the heat transport path through electrical isolation of the modules. The module used in this study consisted of a beryllium core, and electrical cartridge heater simulating the TFE, and three heat pipes to dissipate the waste heat. The investigation was focused on the thermal performance of the assembly, including evaluation of the sodium and braze bonding options for minimizing the thermal resistance between the elements, the temperature distribution in the beryllium matrix, and the heat pipe performance. Continuing subjects of the investigation include performance of the heat pipes through start-up transients, during normal operation, and in a single heat pipe failure mode. Secondary objectives of the investigation include correlation of analytic models for the thermionic element and module including the effects of gap thermal conductances at the modules electrically insulated surfaces.

  10. Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, David W.; Nielsen, Joseph W.; Norman, Daren R.

    2017-07-01

    The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be well outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.

  11. ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis; Gordon Kohse; Yakov Ostrovsky; David M. Carpenter; Joy L. Rempe

    2015-04-01

    A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Current Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in

  12. Production test PTA-002, increased graphite temperature limit -- B, C and D Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Russell, A.

    1965-12-17

    The fundamental objective of the graphite temperature limit is to prevent excessive oxidation of the graphite moderator blocks with carbon dioxide and water vapor in the reactor atmosphere. Laboratory tests have shown that 10% uniform oxidation of graphite results in a loss in strength of approximately 50%. Production Test IP-725 was conducted at F Reactor for a period of six months at graphite temperatures approximately 50 and 100 C higher than the present graphite temperature limit of 650 C. The results from the F Reactor test suggest that an increase in the graphite temperature limit from 650 C to 700 C is technically feasible from the standpoint of oxidation of the graphite moderator with CO{sub 2}. Any significant additional increase was shown to lead to excessively high oxidation rates and is therefore not considered feasible. The objective of this test, therefore, is to extend the higher temperature investigations to B, C, and D Reactors. For the duration of this test, the graphite temperature limit will be increased from 650 C and 700 C, corresponding to an increase in the graphite stringer temperature limit from 735 C to 790 C. The test is expected to last for approximately six months but may be terminated early on any or all the reactors.

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

    Science.gov (United States)

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

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

  14. Subsize specimen testing of nuclear reactor pressure vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A.S. [Missouri Univ., Rolla, MO (United States). Materials Research Center; Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States); Cannon, N.S. [Westinghouse Hanford Co., Richland, WA (United States); Hamilton, M.L. [Pacific Northwest Lab., Richland, WA (United States)

    1991-12-31

    A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, A533B. The methodology appears to be more satisfactory than the methodologies proposed earlier. USE of a notched-only specimen is partitioned into macro-crack initiation and crack propagation energies. USE of a notched and precracked specimen provides the crack propagation energy. {Delta}USE, the difference between the USE`s of notched-only and precracked specimens, is an estimate of the crack initiation energy. {Delta}USE was normalized by a factor involving the dimensions of the Charpy specimen and the stress concentration factor at the notch root. The normalized values of the {Delta}USE were found to be invariant with specimen size.

  15. Subsize specimen testing of nuclear reactor pressure vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A.S. (Missouri Univ., Rolla, MO (United States). Materials Research Center); Rosinski, S.T. (Sandia National Labs., Albuquerque, NM (United States)); Cannon, N.S. (Westinghouse Hanford Co., Richland, WA (United States)); Hamilton, M.L. (Pacific Northwest Lab., Richland, WA (United States))

    1991-01-01

    A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, A533B. The methodology appears to be more satisfactory than the methodologies proposed earlier. USE of a notched-only specimen is partitioned into macro-crack initiation and crack propagation energies. USE of a notched and precracked specimen provides the crack propagation energy. [Delta]USE, the difference between the USE's of notched-only and precracked specimens, is an estimate of the crack initiation energy. [Delta]USE was normalized by a factor involving the dimensions of the Charpy specimen and the stress concentration factor at the notch root. The normalized values of the [Delta]USE were found to be invariant with specimen size.

  16. Limitations of eddy current testing in a fast reactor environment

    Science.gov (United States)

    Wu, Tao; Bowler, John R.

    2016-02-01

    The feasibility of using eddy current probes for detecting flaws in fast nuclear reactor structures has been investigated with the aim of detecting defects immersed in electrically conductive coolant including under liquid sodium during standby. For the inspections to be viable, there is a need to use an encapsulated sensor system that can be move into position with the aid of visualization tools. The initial objective being to locate the surface to be investigated using, for example, a combination of electromagnetic sensors and sonar. Here we focus on one feature of the task in which eddy current probe impedance variations due to interaction with the external surface of a tube are evaluated in order to monitor the probe location and orientation during inspection.

  17. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    Energy Technology Data Exchange (ETDEWEB)

    Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W. [and others

    1995-08-01

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program.

  18. Sequential probability ratio tests for reactor signal validation and sensor surveillance applications

    Energy Technology Data Exchange (ETDEWEB)

    Humenik, K. (Maryland Univ., Baltimore, MD (USA)); Gross, K.C. (Argonne National Lab., IL (USA))

    1989-11-09

    This paper examines the properties of sequential probability ratio tests (SPRT's) and the application of these tests to nuclear power reactor operation. Recently SPRT's have been applied to delayed-neutron (DN) signal data analysis using actual reactor data from the Experimental Breeder Reactor-II, which is operated by Argonne National Laboratory. The implementation of this research as part of an expert system is described. Mathematical properties of the SPRT are investigated, and theoretical results are validated with tests that use DN-signal data taken from the EBR-II in Idaho. Variations of the basic SPRT and applications to general signal validation are also explored. 16 refs., 3 figs.

  19. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    Energy Technology Data Exchange (ETDEWEB)

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  20. Testing and qualification of Control and Safety Rod and its drive mechanism of Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rajan Babu, V., E-mail: vrb@igcar.gov.i [Indira Gandhi Centre for Atomic Research, Department of Atomic Energy, Kalpakkam 603 102 (India); Veerasamy, R.; Patri, Sudheer; Ignatius Sundar Raj, S.; Kumar Krovvidi, S.C.S.P.; Dash, S.K.; Meikandamurthy, C.; Rajan, K.K.; Puthiyavinayagam, P.; Chellapandi, P.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Department of Atomic Energy, Kalpakkam 603 102 (India)

    2010-07-15

    Prototype Fast Breeder Reactor (PFBR) has two independent fast acting diverse shutdown systems. The absorber rod of the first system is called Control and Safety Rod (CSR). CSR and its Drive Mechanism (CSRDM) are used for reactor control and for safe shutdown of the reactor by scram action. In view of the safety role, the qualification of CSRDM is one of the important requirements. CSR and CSRDM were qualified in two stages by extensive testing. In the first stage, the critical subassemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot and dynamic seals and CSR subassembly, were tested and qualified individually simulating the operating conditions of the reactor. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts of the mechanism. In the second stage, full-scale CSRDM and CSR were subjected to all the integrated functional tests in air, hot argon and subsequently in sodium simulating the operating conditions of the reactor and finally subjected to endurance tests. Since the damage occurring in CSRDM and CSR is mainly due to fatigue cycles during scram actions, the number of test cycles was decided based on the guidelines given in ASME, Section III, Div. 1. The results show that the performance of CSRDM and CSR is satisfactory. Subsequent to the testing in sodium, the assemblies having contact with liquid sodium/sodium vapour were cleaned using CO{sub 2} process and the total cleaning process has been established, so that the mechanism can be reused in sodium. The various stages of qualification programmes have raised the confidence level on the performance of the system as a whole for the intended and reliable operation in the reactor.

  1. Design of Seismic Test Rig for Control Rod Drive Mechanism of Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Jongoh; Kim, Gyeongho; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The reactor assembly is submerged in a reactor pool filled with water and its reactivity is controlled by locations of four control absorber rods(CARs) inside the reactor assembly. Each CAR is driven by a stepping motor installed at the top of the reactor pool and they are connected to each other by a tie rod and an electromagnet. The CARs scram the reactor by de-energizing the electromagnet in the event of a safe shutdown earthquake(SSE). Therefore, the safety function of the control rod drive mechanism(CRDM) which consists of a drive assembly, tie rod and CARs is to drop the CAR into the core within an appropriate time in case of the SSE. As well known, the operability for complex equipment such as the CRDM during an earthquake is very hard to be demonstrated by analysis and should be verified through tests. One of them simulates the reactor assembly and the guide tube of the CAR, and the other one does the pool wall where the drive assembly is installed. In this paper, design of the latter test rig and how the test is performed are presented. Initial design of the seismic test rig and excitation table had its first natural frequency at 16.3Hz and could not represent the environment where the CRDM was installed. Therefore, experimental modal analyses were performed and an FE model for the test rig and table was obtained and tuned based on the experimental results. Using the FE model, the design of the test rig and table was modified in order to have higher natural frequency than the cutoff frequency. The goal was achieved by changing its center of gravity and the stiffness of its sliding bearings.

  2. Cold Model Study and Commercial Test on Novel Vapor-Liquid Distributor of Hydroprocessing Reactor

    Institute of Scientific and Technical Information of China (English)

    Wang Shaobing; Zhang Zhanzhu; Wu Defei; Guo Qingming

    2007-01-01

    A novel vapor-liquid distributor was developed on the basis of sufficient study on the existing distributors applied in hydroprocessing reactors.The cold model test data showed that the fluid distribution performance of the novel vapor-liquid distributor was evidently better than the traditional one.Commercial tests of the new distributor were carried out in the 300 kt/a gas oil hydrotreating reactor at SINOPEC Changling Branch Company,showing that the new vapor-liquid distributor could improve the fluid distribution,promote the hydrotreating efficiency and lead to better performance than the traditional one.

  3. Postirradiation examination of recycle test elements from the Peach Bottom Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tiegs, T.N.; Long, E.L. Jr.

    1978-12-01

    The Recycle Test Elements were a series of tests of High-Temperature Gas-Cooled Reactor fuels irradiated in Core 2 of the Peach Bottom Unit 1 Reactor. They tested a wide variety of fissile and fertile fuel types of prime interest when the tests were designed. The fuel types included UO/sub 2/, UC/sub 2/, (2Th,U)O/sub 2/, (4Th,U)O/sub 2/, ThC/sub 2/, and ThO/sub 2/. The mixed thorium--uranium oxides and the pure thorium oxide were tested as Biso-coated particles only, while the others were tested as both Biso- and Triso-coated particles. The Biso coatings on the fissile kernels contained the fission products inadequately but on the fertile kernels they did so acceptably. The results from accelerated and real-time tests on the particle types agreed well.

  4. Development of Digitex premier digital angiographic systems

    Energy Technology Data Exchange (ETDEWEB)

    Miura, Yoshiaki; Miura, Yusuke; Goto, Keiichi; Imanishi, Tetsuo; Miyamoto, Wataru [Shimadzu Corp., Medical Systems Division, Kyoto (Japan)

    2003-06-01

    The technique of interventional radiology has come to be widely utilized in the field of angiography. This has brought forth a strong demand that digital angiographic systems provide high efficiency in patient examinations and high level of interventional support. This report refers to our newly developed Digitex Premier Series digital angiographic systems, designed to meet the above demands. The new systems utilize a high-speed, wide-range C-arm system, a high-resolution image intensifier, a fluid-lubricant X-ray tube, and a digital image processing system, in order to ensure high patient examination efficiency. Their IVR (interventional radiology)-Master bed-side image controller further enhances the efficiency of patient examinations, and also, their CAT (comfortable angio terminal) and FMC (file management console) improve the patient examination throughput and diagnostic workflow of the systems. (author)

  5. Fusion Reactor and Fusion Reactor Materials:Concept Design of the ITER Test Blanket Modules

    Institute of Scientific and Technical Information of China (English)

    HUANGJinhua; LIZaixing; ZHUYukun; HUGang

    2003-01-01

    Performances required: prospect to be adopted in DEMO. Shielding for V.V. and TFC in ITER. Design principles: the peak temperature and stress should not exceed technical limits. The structure of test blanket modules (TBM) should be simple for easy fabrication, and TBM should be robust for reliability.

  6. Advanced high-pressure bench-scale reactor for testing with hot corrosive gases

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, J.; Bachta, R.P.; Wangerow, J.R. (Inst. of Gas Technology, Chicago, IL (United States)); Mojtahedi, W.; Salo, K. (Enviropower Inc., Espoo (Finland))

    1994-01-01

    A bench-scale, high-pressure/high-temperature fluidized-bed reactor (HPTR) system is described that is capable of operating at a maximum temperature and pressure of 1,000 C and 30 bar in a corrosive atmosphere. The design of the unit is based on a double-shell balanced-pressure system. All the hot parts of the reactor that are wetted by the corrosive (and/or reactive) gases and the entire sampling line are constructed of inert material to prevent corrosion and loss of the reactant gases. The unit has been used for over 200 high-pressure hot coal gas desulfurization tests at 20 bars and up to 750 C without any experimental problem and with excellent sulfur balance, indicating that this reactor system is ideal for testing with reactive and corrosive gases at elevated pressures and temperatures.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  8. High Temperature Stress Analysis on 61-pin Test Assembly for Reactor Core Sub-channel Flow Test

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dongwon; Kim, Hyungmo; Lee, Hyeongyeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this study, a high temperature heat transfer and stress analysis of a 61-pin test fuel assembly scaled down from the full scale 217-pin sub-assembly was conducted. The reactor core subchannel flow characteristic test will be conducted to evaluate uncertainties in computer codes used for reactor core thermal hydraulic design. Stress analysis for a 61-pin fuel assembly scaled down from Prototype Generation IV Sodium-cooled Fast Reactor was conducted and structural integrity in terms of load controlled stress limits was conducted. In this study, The evaluations on load-controlled stress limits for a 61-pin test fuel assembly to be used for reactor core subchannel flow distribution tests were conducted assuming that the test assembly is installed in a Prototype Generation IV Sodium-cooled fast reactor core. The 61-pin test assembly has the geometric similarity on P/D and H/D with PGSFR and material of fuel assembly is austenitic stainless steel 316L. The stress analysis results showed that 4.05MPa under primary load occurred at mid part of the test assembly and it was shown that the value of 4.05Mpa was far smaller than the code allowable of 127MPa. , it was shown that the stress intensity due to due to primary load is very small. The stress analysis results under primary and secondary loads showed that maximum stress intensity of 84.08MPa occurred at upper flange tangent to outer casing and the value was well within the code allowable of 268.8MPa. Integrity evaluations based on strain limits and creep-fatigue damage are underway according to the elevated design codes.

  9. Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rule, K.; Scott, J.; Larson, S. [Princeton Plasma Physics Lab., NJ (United States)] [and others

    1995-12-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methods for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.

  10. Performance tests of a small hydrogen reactor based on Mg-Al pellets

    Energy Technology Data Exchange (ETDEWEB)

    Capurso, Giovanni, E-mail: giovanni.capurso@studenti.unipd.it [Dipartimento di Ingegneria Meccanica, Settore Materiali, Universita di Padova, via Marzolo 9, 35131 Padova (Italy); Agresti, Filippo [Dipartimento di Ingegneria Meccanica, Settore Materiali, Universita di Padova, via Marzolo 9, 35131 Padova (Italy); Russo, Sergio Lo [Dipartimento di Fisica and CNISM, Universita di Padova, via Marzolo 8, 35131 Padova (Italy); Maddalena, Amedeo; Principi, Giovanni [Dipartimento di Ingegneria Meccanica, Settore Materiali, Universita di Padova, via Marzolo 9, 35131 Padova (Italy); Cavallari, Andrea; Guardamagna, Cristina [ERSE s.p.a., via Rubattino 54, 20134 Milano (Italy)

    2011-09-15

    On the basis of a previously acquired experience on scaling up issues concerning the use of magnesium hydride as a base material for solid-state hydrogen storage, a small reactor was designed and tested in different operating conditions. It contains about 10 g of catalyzed magnesium hydride powder mixed with 5 wt.% aluminium powder and pressed in the form of cylindrical pellets and the heat flow is managed by means of an oil circulation system. Carbon paper is used to ensure good heat conductivity between the pellets and the inner wall of the reactor and between one pellet and another. A number of hydrogen absorption and desorption cycles at different temperatures and pressures was carried out to compare the behaviour of the small reactor with the laboratory data obtained on small amounts (fractions of grams) of powdered and pelletized samples. Data acquisition for gas flow, pressure and temperature in different positions of the reactor allow a good understanding of internal dynamics. The results in terms of hydrogen absorption/desorption kinetics and of stability to ongoing cycles are stimulating, so that the tested small reactor can be considered as a basic element for further studies and improvements.

  11. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY1999-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    The HTTR (High Temperature Engineering Test Reactor) with the thermal power of 30 MW and the reactor outlet coolant temperature of 850/950 degC is the first high temperature gas-cooled reactor (HTGR) in Japan, which uses coated fuel particle, graphite for core components, and helium gas for primary coolant. The HTTR, which locates at the south-west area of 50,000 m{sup 2} in the Oarai Research Establishment, had been constructed since 1991 before accomplishing the first criticality on November 10, 1998. Rise to power tests of the HTTR started in September, 1999 and the rated thermal power of 30 MW and the reactor outlet coolant temperature of 850 degC was attained in December 2001. JAERI received the certificate of pre-operation test, that is, the commissioning license for the HTTR in March 2002. This report summarizes operation, tests, maintenance, radiation control, and construction of components and facilities for the HTTR as well as R and Ds on HTGRs from FY1999 to 2001. (author)

  12. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-10-01

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition.

  13. Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests

    Energy Technology Data Exchange (ETDEWEB)

    Braley, Jenifer C.; Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-06-15

    Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

  14. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2010-01-01

    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

  15. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, N.J.

    1998-10-21

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints.

  16. Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1996-08-01

    This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR.

  17. Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D.D.; Collins, J.L.

    2000-02-01

    One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.

  18. Digital System Reliability Test for the Evaluation of safety Critical Software of Digital Reactor Protection System

    Directory of Open Access Journals (Sweden)

    Hyun-Kook Shin

    2006-08-01

    Full Text Available A new Digital Reactor Protection System (DRPS based on VME bus Single Board Computer has been developed by KOPEC to prevent software Common Mode Failure(CMF inside digital system. The new DRPS has been proved to be an effective digital safety system to prevent CMF by Defense-in-Depth and Diversity (DID&D analysis. However, for practical use in Nuclear Power Plants, the performance test and the reliability test are essential for the digital system qualification. In this study, a single channel of DRPS prototype has been manufactured for the evaluation of DRPS capabilities. The integrated functional tests are performed and the system reliability is analyzed and tested. The results of reliability test show that the application software of DRPS has a very high reliability compared with the analog reactor protection systems.

  19. Mechanical properties of reactor pressure vessel steels studied by static and dynamic torsion tests

    Science.gov (United States)

    Munier, A.; Maamouri, M.; Schaller, R.; Mercier, O.

    1993-06-01

    Internal friction measurements and torsional plastic deformation tests have been performed in reactor pressure vessel steels (unirradiated, irradiated and irradiated/annealed specimens). The results of these experiments have been interpreted with help of transmission electron microscopy observations (conventional and in situ). It is shown how the interactions between screw dislocations and obstacles (Peierls valleys, impurities and precipitates) could explain the low temperature hardening and the irradiation embrittlement of ferritic steels. In addition, it appears that the nondestructive internal friction technique could be used advantageously to follow the evolution of the material properties under irradiation, as for instance the irradiation embrittlement of the reactor pressure vessel steels.

  20. Acoustic emission monitoring of hot functional testing: Watts Bar Unit 1 Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hutton, P.H.; Dawson, J.F.; Friesel, M.A.; Harris, J.C.; Pappas, R.A.

    1984-06-01

    Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Power Plant during hot functional preservice testing is described in this report. The report deals with background, methodology, and results. The work discussed here is a major milestone in a program supported by NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing toward AE monitoring during reactor operation.

  1. Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine

    Energy Technology Data Exchange (ETDEWEB)

    Reilly, Raymond W.

    2012-07-30

    This project, Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine was established at the Kharkiv Institute of Physics and Technology (KIPT). The associated CRADA was established with Campbell Applied Physics (CAP) located in El Dorado Hills, California. This project extends an earlier project involving both CAP and KIPT conducted under a separate CRADA. The initial project developed the basic Plasma Chemical Reactor (PCR) for generation of ozone gas. This project built upon the technology developed in the first project, greatly enhancing the output of the PCR while also improving reliability and system control.

  2. Pilot-plant testing of magnetic filters for the N-Reactor primary cooling circuit

    Energy Technology Data Exchange (ETDEWEB)

    Emory, B.B.

    1982-01-01

    Data obtained during the laboratory loop test program using the high power HGMF indicates that removal efficiency for /sup 60/Co and subsequently the bulk of the crud, will be greater than 90% at field strength above .1 Tesla for the expanded metal mesh matrix. However, since /sup 54/Mn seems to exhibit paramagnetic behavior and the possibility of quantities of alpha iron forming during reactor shut down from oxygen inleakage, a field strength of .5 to 1 Tesla may be more appropriate for a full scale on-reactor installation. Crud loading of 50 gm per kg of matrix weight are readily obtainable and up to twice that amount has been reached.

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

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

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

  4. Tests of Lorentz and CPT Violation in the Medium Baseline Reactor Antineutrino Experiment

    CERN Document Server

    Li, Yu-Feng

    2014-01-01

    Tests of Lorentz and CPT violation in the medium baseline reactor antineutrino experiment are presented in the framework of the Standard Model Extension (SME). Both the spectral distortion and sidereal variation are employed to derive the limits of Lorentz violation (LV) coefficients. We do the numerical analysis of the sensitivity of LV coefficients by taking the Jiangmen Underground Neutrino Observatory (JUNO) as an illustration, which can improve the sensitivity by more than two orders of magnitude compared with the current limits from reactor antineutrino experiments.

  5. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

  6. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

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

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

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

  8. Evaluation of the Shielding Characteristics Test around the Reactor Core in the Prototype Fbr Monju

    Science.gov (United States)

    Usami, Shin; Suzuoki, Zenro; Deshimaru, Takehide; Nakashima, Fumiaki; Hikichi, Takuo

    2003-06-01

    In Monju, shielding measurements were made around the reactor core as a part of the system start-up tests in order to evaluate the design margins of the shielding performance, to demonstrate the validity of the shielding analysis method, and to acquire basic data for use in future FBR design. The measured reaction rates have been obtained radially from the core to the in-vessel storage rack and axially to the reactor vessel upper plenum. The measured values (E) were compared with the calculated values (C) obtained with the FBR shielding analysis system on the basis of the nuclear data library JENDL-3.2. Based upon these results, the design margins around the reactor core have been examined.

  9. Two-Compartment Photoelectrochemical Reactors Tested under various solar Light Concentration ratios

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez-Ibanez, P.; Malato, S. [Plataforma Solar de Almeria. CIEMAT (Spain)

    1999-07-01

    A new type of photo reactor, made of a cylindrical photo anode placed around an inner comportment, has been adapted to the compound parabolic (CPC's) and to the parabolic trough (PTC, Helioman) solar collectors. The photoelectrochemical performances of such two-compartment photo reactors are noticeably improved with respect to those previously obtained with photo reactors having flat photoanodes and a single compartment. The abatement of model pollutants shows up to threefold higher organic oxidation rates compared to Ti O{sub 2} slurries tested in the same experimental conditions. Clearly, charge separation is much better when an external electrochemical bias is applied to Ti/Ti O{sub 2} photoanodes under irradiation. (Author) 8 refs.

  10. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Vivek [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Human Factors, Controls, and Statistics; Smith, James A. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design; Jewell, James Keith [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design

    2015-02-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  11. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Cooling Systems for New Boiling-Water Reactors.'' This RG describes testing methods the NRC staff considers acceptable for demonstrating the operability of emergency core cooling systems (ECCSs) for boiling...

  12. Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

    Science.gov (United States)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

    The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

  13. The RES reactor, a test reactor for naval propulsion; Le reacteur d'essais RES, reacteur d'essais de la propulsion navale

    Energy Technology Data Exchange (ETDEWEB)

    Pivet, S. [CEA Bruyeres-le-Chatel, 91 (France); Minguet, J.L. [AREVA-Technicatome, 13 - Aix en Provence (France)

    2005-07-01

    The RES, the new test reactor for naval propulsion, will replace the RNG that nears the end of its operating life after 30 years in service. The main asset of a land-based installation is to provide an in-core instrumented reactor while the on-board system must stay as simple as possible for robustness reasons. The objective of the RES is fivefold: 1) to foresee and help solving problems likely to happen on on-board reactor, 2) to validate nuclear fuels and reactor systems for naval propulsion, 3) to validate reactor system and equipment for the Barracuda submarine program, 4) to upgrade the on-ground facility located at Cadarache, and 5) to provide the Cea with a new capacity for the storing of spent fuels from naval propulsion systems and from Cea research reactors. The RES facility is made of 2 parts: one that houses the reactor and the other that is dedicated to the handling on spent fuels, their examination through a gamma spectrometry bench and their storing in a pool. The RES facility is scheduled to open in 2009. (A.C.)

  14. Preliminary Benchmark Evaluation of Japan’s High Temperature Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John Darrell Bess

    2009-05-01

    A benchmark model of the initial fully-loaded start-up core critical of Japan’s High Temperature Engineering Test Reactor (HTTR) was developed to provide data in support of ongoing validation efforts of the Very High Temperature Reactor Program using publicly available resources. The HTTR is a 30 MWt test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. The benchmark was modeled using MCNP5 with various neutron cross-section libraries. An uncertainty evaluation was performed by perturbing the benchmark model and comparing the resultant eigenvalues. The calculated eigenvalues are approximately 2-3% greater than expected with an uncertainty of ±0.70%. The primary sources of uncertainty are the impurities in the core and reflector graphite. The release of additional HTTR data could effectively reduce the benchmark model uncertainties and bias. Sensitivity of the results to the graphite impurity content might imply that further evaluation of the graphite content could significantly improve calculated results. Proper characterization of graphite for future Next Generation Nuclear Power reactor designs will improve computational modeling capabilities. Current benchmarking activities include evaluation of the annular HTTR cores and assessment of the remaining start-up core physics experiments, including reactivity effects, reactivity coefficient, and reaction-rate distribution measurements. Long term benchmarking goals might include analyses of the hot zero-power critical, rise-to-power tests, and other irradiation, safety, and technical evaluations performed with the HTTR.

  15. The 5th surveillance testing for Kori unit 2 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

  16. The 4th surveillance testing for Kori unit 3 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-10-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

  17. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

    2010-06-01

    2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed

  18. Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

    2012-08-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

  19. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Quiping [The Ohio State Univ., Columbus, OH (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Chtistensen, Richard [The Ohio State Univ., Columbus, OH (United States); Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Yoder, Graydon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  20. Non destructive testing of irradiated fuel assemblies at the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Jose Eduardo Rosa da; Terremoto, Luis Antonio Albiac; Castanheira, Myrthes; Teodoro, Celso Antonio; Silva, Antonio Teixeira e; Damy, Margaret de Almeida; Lucki, Georgi [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: jersilva@ipen.br; laaterre@ipen.br; myrthes@ipen.br; cteodoro@ipen.br; teixeira@ipen.br; madamy@ipen.br; glucki@ipen.br

    2007-07-01

    Fuel performance and nuclear fuel qualification require a post-irradiation analysis. Non-destructive methods are utilised both in irradiated fuel storage pools and in hot-cells laboratories. As Brazil does not have hot-cells facilities for post-irradiation analysis, a qualification program for the Material Testing Reactor (MTR) fuel elements made at IPEN/CNEN-SP was adopted, based on non-destructive tests. The IPEN Fuel Engineering Group - CENC developed basic facilities for fuels post-irradiated analysis inside the reactor pool, which gives indications of: general state, by visual inspection; the integrity of the irradiated fuel cladding, by sipping tests; thickness measurements of the fuel miniplates during the irradiation time, for swelling evaluation; and, local burn-up evaluation by gamma spectrometry along the active area of the fuel element. This work describes that facilities, equipment and examples of some irradiated fuels analysis performed. (author)

  1. The 5th surveillance testing for Kori unit 1 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed primarily by Korea Atomic Energy Research Institute and Westinhouse corporation partially involved in testing and calculation data evaluation in order to obtain reliable test result. Fast neutron fluences for capsule V, T, S, R and P were 5.087E+18, 1.115E+19, 1.228E+19, 2.988E+19, and 3.938E+19n/cm2, respectively. The bias factor, the ratio of calculation/measurement, was 0.940 for the 1st through 5th testing and the calculational uncertainty, 7% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.9846E+19n/cm{sup 2} based on the end of 17th fuel cycle and it was predicted that the fluences of vessel inside surface at 24, 32, 40 and 48EFPY would reach 3.0593E+19, 4.0695E+19, 5.0797E+19 and 6.0900E+19n/cm{sup 2} based on the current calculation. PTS analysis for Kori unit 1 showed that 27.93EFPY was the threshold value for 300 deg F requirement. 71 refs., 33 figs., 52 tabs. (Author)

  2. In-reactor tests of the nuclear light bulb rocket concept

    Science.gov (United States)

    Gauntt, R. O.; Slutz, S. A.; Latham, T. S.; Roman, W. C.; Rogers, R. J.

    1992-07-01

    An overview is given of the closed-cycle Gas Core Nuclear Rocket outlining scenarios for its use in short-duration Mars missions and results of Nuclear Light Bulb (NLB) tests. Isothermal and nonnuclear tests are described which confirmed the fundamental concepts behind the NLB. NLB reference-engine performance characteristics are given for hypothetical engines that could be used for manned Mars missions. Vehicle/propulsion sizing is based on a Mars mission with three trans-Mars impulse burns, capture and escape burns, and a total mission duration of 600 days. The engine would have a specific impulse of 1870 seconds, a 412-kN thrust, and a thrust/weight ratio of 1.3. Reactor tests including small-scale in-reactor tests are shown to be prerequisites for studying: (1) fluid mechanical confinement of the gaseous nuclear fuel; (2) buffer gas separation and circulation; and (3) the minimization of transparent wall-heat loading. The reactor tests are shown to be critical for establishing the feasibility of the NLB concept.

  3. Proving test on the seismic reliability of nuclear power plant: PWR reactor containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi; Yoshikawa, Teiichi; Ohno, Tokue; Yoshikawa, Eiji.

    1989-01-01

    Seismic reliability proving tests of nuclear power plant facilities are carried out by the Nuclear Power Engineering Test Center, using the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry. In 1982, the seismic reliability proving test of a PWR containment vessel was conducted using a test component of reduced scale 1/3.7. As a result of this test, the test component proved to have structural soundness against earthquakes, and at the same time its stable function was proved by leak tests which were carried out before and after the vibration test. In 1983, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. The seismic analysis and evaluation on the actual containment vessel were then performed using these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed.

  4. Assessment of Startup Fuel Options for a Test or Demonstration Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, Jon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Walters, L. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO2 and UO2-PuO2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availability are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.

  5. Criticality Safety Evaluation for the Advanced Test Reactor U-Mo Demonstration Elements

    Energy Technology Data Exchange (ETDEWEB)

    Leland M. Montierth

    2010-12-01

    The Reduced Enrichment Research Test Reactors (RERTR) fuel development program is developing a high uranium density fuel based on a (LEU) uranium-molybdenum alloy. Testing of prototypic RERTR fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. Two RERTR-Full Size Demonstration fuel elements based on the ATR-Reduced YA elements (all but one plate fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). The two fuel elements will be irradiated in alternating cycles such that only one element is loaded in the reactor at a time. Existing criticality analyses have analyzed Standard (HEU) ATR elements (all plates fueled) from which controls have been derived. This criticality safety evaluation (CSE) documents analysis that determines the reactivity of the Demonstration fuel elements relative to HEU ATR elements and shows that the Demonstration elements are bound by the Standard HEU ATR elements and existing HEU ATR element controls are applicable to the Demonstration elements.

  6. Le diagnostic anténatal de la trisomie 21 par l'hybridation in situ en fluorescence (FISH): à propos des premiers tests réalisés au Maroc

    Science.gov (United States)

    Lamzouri, Afaf; Natiq, Abdelhafid; Tajir, Mariam; Sendid, Mohamed; Sefiani, Abdelaziz

    2012-01-01

    Introduction Le but de cette étude était de présenter les premiers résultats de diagnostic anténatal de la trisomie 21 par la technique d'hybridation in situ en fluorescence (FISH) au Maroc et discuter son intérêt dans le diagnostic rapide de cette aneuploïdie. Méthodes Ce travail a été réalisé chez 23 femmes avec des grossesses à haut risque de trisomie 21. La moyenne d’âge des gestantes étaient de 37,43 ans avec des extrêmes de 21 et 43 ans. Toutes étaient musulmanes mariées, mariage légitimé par la Charia, dont trois mariages consanguins, sauf une originaire de la République Démocratique du Congo qui était chrétienne et concubine. La majorité des femmes étaient fonctionnaires et avaient un niveau de scolarisation moyen à élevé. Toutes les patientes ont bénéficié d'une consultation de génétique médicale au cours de laquelle il leur a été donné des informations sur la technique, son intérêt et ses limites. Il s'agit de femmes enceintes qui avaient soit un âge maternel élevé ou des signes d'appel échographiques et/ ou biochimiques. Une des patientes était porteuse d'une translocation robertsonienne t(14;21) équilibrée. Une amniocentèse a été réalisée chez toutes les gestantes et aucun avortement n'a était induit par ce geste invasif. L’âge gestationnel moyen à la première consultation était de 14 semaines d'aménorrhée (SA) et à l'amniocentèse était de 16 SA et 5 jours. L'analyse FISH a été réalisée, après consentement des couples, sur des cellules non cultivées à partir des échantillons de liquides amniotiques, en utilisant des sondes spécifiques du chromosome 21. Résultats Parmi les 23 patientes qui ont bénéficiées d'un diagnostic anténatal de la trisomie 21 par la technique FISH, nous avons pu rassurer 21 d'entre elles, et nous avons détecté deux cas de trisomie 21 fœtal. Conclusion La technique FISH permet un diagnostic anténatal rapide, en moins de 48h, de la trisomie 21 sur

  7. Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)

    2012-02-01

    Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental

  8. Metaphor Analysis of Chinese Premier Wen’s Cambridge Speech

    Institute of Scientific and Technical Information of China (English)

    LUO Luo

    2014-01-01

    Metaphor is more than an ostensible decoration of language. It is an integral part of human thought of ideologized world. This article analyzes the metaphor use of Chinese Premier Wen Jiabao’s speech at Cambridge in February 2009, in an at-tempt to display how the preferred metaphors serve the purpose of this speech and reflect Premier Wen ’s construction of Chi-na’s situation.

  9. Feasibility of conducting a dynamic helium charging experiment for vanadium alloys in the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.; Strain, R.V.; Smith, D.L. [Argonne National Lab., IL (United States); Matsui, H. [Tohoku Univ. (Japan)

    1996-10-01

    The feasibility of conducting a dynamic helium charging experiment (DHCE) for vanadium alloys in the water-cooled Advanced Test Reactor (ATR) is being investigated as part of the U.S./Monbusho collaboration. Preliminary findings suggest that such an experiment is feasible, with certain constraints. Creating a suitable irradiation position in the ATR, designing an effective thermal neutron filter, incorporating thermocouples for limited specimen temperature monitoring, and handling of tritium during various phases of the assembly and reactor operation all appear to be feasible. An issue that would require special attention, however, is tritium permeation loss through the capsule wall at the higher design temperatures (>{approx}600{degrees}C). If permeation is excessive, the reduced amount of tritium entering the test specimens would limit the helium generation rates in them. At the lower design temperatures (<{approx}425{degrees}C), sodium, instead of lithium, may have to be used as the bond material to overcome the tritium solubility limitation.

  10. Field test and evaluation of the passive neutron coincidence collar for prototype fast reactor fuel subassemblies

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, H.O.; Keddar, A.

    1982-08-01

    The passive neutron Coincidence Collar, which was developed for the verification of plutonium content in fast reactor fuel subassemblies, has been field tested using Prototype Fast Reactor fuel. For passive applications, the system measures the /sup 240/Pu-effective mass from the spontaneous fission rate, and in addition, a self-interrogation technique is used to determine the fissile content in the subassembly. Both the passive and active modes were evaluated at the Windscale Works in the United Kingdom. The results of the tests gave a standard deviation 0.75% for the passive count and 3 to 7% for the active measurement for a 1000-s counting time. The unit will be used in the future for the verification of plutonium in fresh fuel assemblies.

  11. A preliminary neutronic evaluation of high temperature engineering test reactor using the SCALE6 code

    Science.gov (United States)

    Tanure, L. P. A. R.; Sousa, R. V.; Costa, D. F.; Cardoso, F.; Veloso, M. A. F.; Pereira, C.

    2014-02-01

    Neutronic parameters of some fourth generation nuclear reactors have been investigated at the Departamento de Engenharia Nuclear/UFMG. Previous studies show the possibility to increase the transmutation capabilities of these fourth generation systems to achieve significant reduction concerning transuranic elements in spent fuel. To validate the studies, a benchmark on core physics analysis, related to initial testing of the High Temperature Engineering Test Reactor and provided by International Atomic Energy Agency (IAEA) was simulated using the Standardized Computer Analysis for Licensing Evaluation (SCALE). The CSAS6/KENO-VI control sequence and the 44-group ENDF/B-V 0 cross-section neutron library were used to evaluate the keff (effective multiplication factor) and the result presents good agreement with experimental value.

  12. Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-03-01

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

  14. After Action Report: Advanced Test Reactor Complex 2015 Evaluated Drill October 6, 2015

    Energy Technology Data Exchange (ETDEWEB)

    Holmes, Forest Howard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-11-01

    The Advanced Test Reactor (ATR) Complex, operated by Battelle Energy Alliance, LLC, at the Idaho National Laboratory (INL) conducted an evaluated drill on October 6, 2015, to allow the ATR Complex emergency response organization (ERO) to demonstrate the ability to respond to and mitigate an emergency by implementing the requirements of DOE O 151.1C, “Comprehensive Emergency Management System.”

  15. Dynamic parameters test of Haiyang Nuclear Power Engineering in reactor areas, China

    Science.gov (United States)

    Zhou, N.; Zhao, S.; Sun, L.

    2012-12-01

    Haiyang Nuclear Power Project is located in Haiyang city, China. It consists of 6×1000MW AP1000 Nuclear Power generator sets. The dynamic parameters of the rockmass are essential for the design of the nuclear power plant. No.1 and No.2 reactor area are taken as research target in this paper. Sonic logging, single hole and cross-hole wave velocity are carried out respectively on the site. There are four types of rock lithology within the measured depth. They are siltstone, fine sandstone, shale and allgovite. The total depth of sonic logging is 409.8m and 2049 test points. The sound wave velocity of the rocks are respectively 5521 m/s, 5576m/s, 5318 m/s and 5576 m/s. Accroding to the statistic data, among medium weathered fine sandstone, fairly broken is majority, broken and relatively integrity are second, part of integrity. Medium weathered siltstone, relatively integrity is mojority, fairly broken is second. Medium weathered shale, fairly broken is majority, broken and relatively integrity for the next and part of integrity. Slight weathered fine sandstone, siltstone, shale and allgovite, integrity is the mojority, relatively integrity for the next, part of fairly broken.The single hole wave velocity tests are set in two boreholesin No.1 reactor area and No.2 reactor area respectively. The test depths of two holes are 2-24m, and the others are 2-40m. The wave velocity data are calculated at different depth in each holes and dynamic parameters. According to the test statistic data, the wave velocity and the dynamic parameter values of rockmass are distinctly influenced by the weathering degree. The test results are list in table 1. 3 groups of cross hole wave velocity tests are set for No.1 and 2 reactor area, No.1 reactor area: B16, B16-1, B20(Direction:175°, depth: 100m); B10, B10-1, B11(269°, 40m); B21, B21-1, B17(154°, 40m); with HB16, HB10, HB21 as trigger holes; No.2 reactor area: B47, B47-1, HB51(176°, 100m); B40, B40-1, B41(272°, 40m); B42, B42-1, B

  16. Highly Perturbed Operational Test Configurations at the WSMR Fast Burst Reactor

    Directory of Open Access Journals (Sweden)

    Flanders T. Michael

    2016-01-01

    Full Text Available The White Sands Missile Range (WSMR MoLLY-G reactor has a long history of producing a well characterized environment for testing electronic systems/devices in fission environments. As an unmoderated, unreflected, bare critical assembly, it provides a slightly degraded fission spectrum with a 1/E tail. For radiation hardness testing of electronics, the neutron fluence is usually reported as the 1-MeV Equivalent Neutron Fluence for Silicon. In this paper, we examine additional neutron environments and characterizations ranging from low intensity neutron fields to more extreme modifications of our normal test environment.

  17. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  18. Initial Testing of the Microscopic Depletion Implementation in the MAMMOTH Reactor Physics Application

    Energy Technology Data Exchange (ETDEWEB)

    J. Ortensi; Y. Wang; S. Schunert; B.D. Ganapol; F.N. Gleicher; B. Baker; M.D. DeHart

    2016-09-01

    Present and new nuclear fuels that will be tested at the Transient Reactor Test (TREAT) facility will be analyzed with the MAMMOTH reactor physics application, currently under development, at Idaho National Laboratory. MAMMOTH natively couples the BISON, RELAP-7, and Rattlesnake applications within the MOOSE framework. This system allows the irradiation of fuel from beginning of life in a nuclear reactor until it is placed in TREAT for fuel testing within the same analysis mesh and, thus, retaining a very high level of resolution and fidelity. The calculation of the isotopic distribution in fuel requires the solution to the decay and transmutation equations coupled to the neutron transport equation. The Chebyshev Rational Approximation Method (CRAM) is the current state-of-the-art in the field, as was chosen to be the solver for the decay and transmutation equations. This report shows that the implementation of the CRAM solver within MAMMOTH is correct with various analytic benchmarks for decay and transmutation of nuclides. The results indicate that the solutions with CRAM order 16 achieve the level of precision of the benchmark. The CRAM solutions show little sensitivity to the time step size and consistently produce a high level of accuracy for isotopic decay for time steps of 1x10^11 years. Comparisons to DRAGON5 with 297 isotopes yield comparable results, but some differences need to be further analyzed.

  19. Long term out-of-pile thermocouple tests in conditions representative for nuclear gas-cooled high temperature reactors

    OpenAIRE

    LAURIE Mathias; FOURREZ Stéphane; FUETTERER Michael; LAPETITE Jean-Marc; SADLI M.; MORICE Ronan; FAILLEAU G

    2013-01-01

    During irradiation tests at high temperature failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. As instrumentation, in particular thermocouples are considered safety-relevant both for irradiation tests and for commercial reactors, JRC and THERMOCOAX joined forces to solve this issue by performing out-of-pile tests with thermocouples mimicking the environment encountered by high temperature reactor (HTR) in-core instrumentation. The objective was to screen innova...

  20. Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, A. J.; Haggard, DC; Herter, J. W.; Swank, W. D.; Knudson, D. L.; Cherry, R. S. [Idaho National Laboratory, P.O. Box 1625, MS 4112, Idaho Falls, ID, (United States); Scervini, M. [University of Cambridge, Department of Material Science and Metallurgy, 27 Charles Babbage Road, CB3 0FS, Cambridge, (United Kingdom)

    2015-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to be only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly

  1. Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

    Energy Technology Data Exchange (ETDEWEB)

    A. J. Palmer; DC Haggard; J. W. Herter; M. Scervini; W. D. Swank; D. L. Knudson; R. S. Cherry

    2011-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina

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

    Energy Technology Data Exchange (ETDEWEB)

    Hakan Ozaltun & Herman Shen

    2011-11-01

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

  3. An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Aaron, Adam M [ORNL; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK); Fugate, David L [ORNL; Holcomb, David Eugene [ORNL; Kisner, Roger A [ORNL; Peretz, Fred J [ORNL; Robb, Kevin R [ORNL; Wilgen, John B [ORNL; Wilson, Dane F [ORNL

    2014-01-01

    The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.

  4. Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

    Science.gov (United States)

    Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

    2017-01-01

    The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

  5. Ultracold neutron source at the PULSTAR reactor: Engineering design and cryogenic testing

    Energy Technology Data Exchange (ETDEWEB)

    Korobkina, E., E-mail: ekorobk@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Box 7909, Raleigh, NC 27695 (United States); Medlin, G. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Wehring, B.; Hawari, A.I. [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Box 7909, Raleigh, NC 27695 (United States); Huffman, P.R.; Young, A.R. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Beaumont, B. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Palmquist, G. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States)

    2014-12-11

    Construction is completed and commissioning is in progress for an ultracold neutron (UCN) source at the PULSTAR reactor on the campus of North Carolina State University. The source utilizes two stages of neutron moderation, one in heavy water at room temperature and the other in solid methane at ∼40K, followed by a converter stage, solid deuterium at 5 K, that allows a single down scattering of cold neutrons to provide UCN. The UCN source rolls into the thermal column enclosure of the PULSTAR reactor, where neutrons will be delivered from a bare face of the reactor core by streaming through a graphite-lined assembly. The source infrastructure, i.e., graphite-lined assembly, heavy-water system, gas handling system, and helium liquefier cooling system, has been tested and all systems operate as predicted. The research program being considered for the PULSTAR UCN source includes the physics of UCN production, fundamental particle physics, and material surface studies of nanolayers containing hydrogen. In the present paper we report details of the engineering and cryogenic design of the facility as well as results of critical commissioning tests without neutrons.

  6. Ultracold neutron source at the PULSTAR reactor: Engineering design and cryogenic testing

    Science.gov (United States)

    Korobkina, E.; Medlin, G.; Wehring, B.; Hawari, A. I.; Huffman, P. R.; Young, A. R.; Beaumont, B.; Palmquist, G.

    2014-12-01

    Construction is completed and commissioning is in progress for an ultracold neutron (UCN) source at the PULSTAR reactor on the campus of North Carolina State University. The source utilizes two stages of neutron moderation, one in heavy water at room temperature and the other in solid methane at ~ 40 K, followed by a converter stage, solid deuterium at 5 K, that allows a single down scattering of cold neutrons to provide UCN. The UCN source rolls into the thermal column enclosure of the PULSTAR reactor, where neutrons will be delivered from a bare face of the reactor core by streaming through a graphite-lined assembly. The source infrastructure, i.e., graphite-lined assembly, heavy-water system, gas handling system, and helium liquefier cooling system, has been tested and all systems operate as predicted. The research program being considered for the PULSTAR UCN source includes the physics of UCN production, fundamental particle physics, and material surface studies of nanolayers containing hydrogen. In the present paper we report details of the engineering and cryogenic design of the facility as well as results of critical commissioning tests without neutrons.

  7. CALMOS: Innovative device for the measurement of nuclear heating in material testing reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carcreff, H. [Alternative Energies and Atomic Energy Commission CEA, Saclay Center, DEN/DANS/DRSN/SIREN, Gif Sur Yvette, 91191 (France)

    2011-07-01

    An R and D program has been carried out since 2002 in order to improve gamma heating measurements in the 70 MWth OSIRIS Material Testing Reactor operated by CEA's Nuclear Energy Div. at the Saclay research center. Throughout this program an innovative calorimetric probe associated to a specific handling system has been designed in order to make measurements both along the fissile height and on the upper part of the core, where nuclear heating rates still remain high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for the process validation, while a displacement system has been especially designed to move the probe axially. A final probe has been designed thanks to modeling results and to preliminary measurements obtained with mock-ups irradiated to a heating level of 2W/g, This paper gives an overview of the development, describes the calorimetric probe, and expected advantages such as the possibility to use complementary methods to get the nuclear heating measurement. Results obtained with mock-ups irradiated in ex-core area of the reactor are presented and discussed. (authors)

  8. Assessment of Feasibility of the Beneficial Use of Waste Heat from the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna P. Guillen

    2012-07-01

    This report investigates the feasibility of using waste heat from the Advanced Test Reactor (ATR). A proposed glycol waste heat recovery system was assessed for technical and economic feasibility. The system under consideration would use waste heat from the ATR secondary coolant system to preheat air for space heating of TRA-670. A tertiary coolant stream would be extracted from the secondary coolant system loop and pumped to a new plate and frame heat exchanger, where heat would be transferred to a glycol loop for preheating outdoor air in the heating and ventilation system. Historical data from Advanced Test Reactor operations over the past 10 years indicates that heat from the reactor coolant was available (when needed for heating) for 43.5% of the year on average. Potential energy cost savings by using the waste heat to preheat intake air is $242K/yr. Technical, safety, and logistics considerations of the glycol waste heat recovery system are outlined. Other opportunities for using waste heat and reducing water usage at ATR are considered.

  9. The development and the tests of the electrostatic probe for dust particle collection in thermonuclear reactors

    Science.gov (United States)

    Begrambekov, L. B.; Voityuk, A. N.; Zakharov, A. M.

    2016-09-01

    Formation of dust particles in thermonuclear reactors can greatly affect the plasma parameters and lead to accumulation of tritium. The rates of formation and deposition of dust need to be measured, and the parameters of formation of dust particles and clusters need to be studied. A model of a device for collection of fine conductive particles capable of removing them from the reactor chamber for future research is proposed in this paper. The dust collector's operation is based on a principle of applied electrostatic field. The model was tested in different operating conditions: in vacuum, at the atmospheric pressure in the atmosphere of air and dry nitrogen. The experiments were conducted with a stationary system and with the dust collector in motion relative to the dusty surface. It is shown that, during the probe moving relative to the surface, it can remove up to 95% of fine tungsten particles with sizes ranging from 1 to 10 μm.

  10. Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    J. Benson; J. Cole; J. Jackson; F. Marshall; D. Ogden; J. Rempe; M. C. Thelen

    2013-02-01

    In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groups conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.

  11. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  12. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    CERN Document Server

    Lumia, M E

    2002-01-01

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-12-31

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

  14. Bayesian Zero-Failure (BAZE) reliability demonstration testing procedure for components of nuclear reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Waller, R.A.

    1977-06-01

    A Bayesian-Zero-Failure (BAZE) reliability demonstration testing procedure is presented. The method is developed for an exponential failure-time model and a gamma prior distribution on the failure-rate. A simple graphical approach using percentiles is used to fit the prior distribution. The procedure is given in an easily applied step-by-step form which does not require the use of a computer for its implementation. The BAZE approach is used to obtain sample test plans for selected components of nuclear reactor safety systems.

  15. Supported Pd-Au Membrane Reactor for Hydrogen Production: Membrane Preparation, Characterization and Testing

    OpenAIRE

    Adolfo Iulianelli; Marjan Alavi; Giuseppe Bagnato; Simona Liguori; Jennifer Wilcox; Mohammad Reza Rahimpour; Reza Eslamlouyan; Bryce Anzelmo; Angelo Basile

    2016-01-01

    A supported Pd-Au (Au 7wt%) membrane was produced by electroless plating deposition. Permeation tests were performed with pure gas (H2, H2, N2, CO2, CH4) for long time operation. After around 400 h under testing, the composite Pd-Au membrane achieved steady state condition, with an H2/N2 ideal selectivity of around 500 at 420 °C and 50 kPa as transmembrane pressure, remaining stable up to 1100 h under operation. Afterwards, the membrane was allocated in a membrane reactor module for methane s...

  16. Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

    Science.gov (United States)

    Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

    2017-08-01

    A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

  17. Test Results From a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

    Science.gov (United States)

    Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.

    2009-01-01

    The Brayton Power Conversion Unit (BPCU) located at NASA Glenn Research Center (GRC) in Cleveland, OH is a closed cycle system incorporating a turboaltemator, recuperator, and gas cooler connected by gas ducts to an external gas heater. For this series of tests, the BPCU was modified by replacing the gas heater with the Direct Drive Gas heater or DOG. The DOG uses electric resistance heaters to simulate a fast spectrum nuclear reactor similar to those proposed for space power applications. The combined system thermal transient behavior was the focus of these tests. The BPCU was operated at various steady state points. At each point it was subjected to transient changes involving shaft rotational speed or DOG electrical input. This paper outlines the changes made to the test unit and describes the testing that took place along with the test results.

  18. Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2012-11-01

    This study investigated the feasibility of using a waste heat recovery system (WHRS) to recover heat from the Advanced Test Reactor (ATR) secondary coolant system (SCS). This heat would be used to preheat air for space heating of the reactor building, thus reducing energy consumption, carbon footprint, and energy costs. Currently, the waste heat from the reactor is rejected to the atmosphere via a four-cell, induced-draft cooling tower. Potential energy and cost savings are 929 kW and $285K/yr. The WHRS would extract a tertiary coolant stream from the SCS loop and pump it to a new plate and frame heat exchanger, from which the heat would be transferred to a glycol loop for preheating outdoor air supplied to the heating and ventilation system. The use of glycol was proposed to avoid the freezing issues that plagued and ultimately caused the failure of a WHRS installed at the ATR in the 1980s. This study assessed the potential installation of a new WHRS for technical, logistical, and economic feasibility.

  19. IMPROVED COMPUTATIONAL NEUTRONICS METHODS AND VALIDATION PROTOCOLS FOR THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Joseph W. Nielsen; Benjamin M. Chase; Ronnie K. Murray; Kevin A. Steuhm

    2012-04-01

    The Idaho National Laboratory (INL) is in the process of modernizing the various reactor physics modeling and simulation tools used to support operation and safety assurance of the Advanced Test Reactor (ATR). Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009 was successfully completed during 2011. This demonstration supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR fuel cycle management process beginning in 2012. On the experimental side of the project, new hardware was fabricated, measurement protocols were finalized, and the first four of six planned physics code validation experiments based on neutron activation spectrometry were conducted at the ATRC facility. Data analysis for the first three experiments, focused on characterization of the neutron spectrum in one of the ATR flux traps, has been completed. The six experiments will ultimately form the basis for a flexible, easily-repeatable ATR physics code validation protocol that is consistent with applicable ASTM standards.

  20. Measurement of basic thermal-hydraulic characteristics under the test facility and reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Eduard A Boltenko; Victor P Sharov [Elektrogorsk Research and Engineering Center, EREC, Bezimyannaja Street, 6, Elektrogorsk, Moscow Region, 142530 (Russian Federation); Dmitriy E Boltenko [State Scientific Center of Russian Federation IPPE, Bondarenko Square, Obhinsk, Kaluga Region, 249020 (Russian Federation)

    2005-07-01

    Full text of publication follows: The nuclear power of Russia is based on the reactors of two types: water-water - WWER and uranium - graphite channel RBMK. The nuclear power development is possible with performance of the basic condition - level of nuclear power plants (NPP) safety should satisfy the rigid requirements. The calculated proof of NPPs safety made by means of thermal-hydraulic codes of improved estimation, verified on experimental data is the characteristic of this level. The data for code verification can be obtained at the integral facilities simulating a circulation circuit of NPP with the basic units and intended for investigation of circuit behaviour in transient and accident conditions. For verification of mathematical models in transient and accident conditions, development of physically reasonable methods for definition of the various characteristics of two-phase flow the experimental data, as the integrated characteristics of a flow, and data on the local characteristics and structure of a flow is necessary. For safety assurance of NPP it is necessary to monitor and determine the basic thermalhydraulic characteristics of reactor facility (RF). It is possible to refer coolant flow-rate, core input and output water temperature, heat-power. The description of the EREC works in the field completion and adaptation of certain methods with reference to measurements in dynamic modes of test facility conditions and development of methods for measurements of basic thermal-hydraulic characteristics of reactor facilities is presented in the paper. (authors)

  1. Bench-scale reactor tests of low-temperature, catalytic gasification of wet, industrial wastes

    Energy Technology Data Exchange (ETDEWEB)

    Elliott, D.C.; Neuenschwander, G.G.; Baker, E.G.; Butner, R.S.; Sealock, L.J.

    1990-04-01

    Bench-scale reactor tests are under way at Pacific Northwest Laboratory to develop a low-temperature, catalytic gasification system. The system, licensed under the trade name Thermochemical Environmental Energy System (TEES{reg sign}), is designed for to a wide variety of feedstocks ranging from dilute organics in water to waste sludges from food processing. The current research program is focused on the use of a continuous-feed, tubular reactor. The catalyst is nickel metal on an inert support. Typical results show that feedstocks such as solutions of 2% para-cresol or 5% and 10% lactose in water or cheese whey can be processed to >99% reduction of chemical oxygen demand (COD) at a rate of up to 2 L/hr. The estimated residence time is less than 5 min at 360{degree}C and 3000 psig, not including 1 to 2 min required in the preheating zone of the reactor. The liquid hourly space velocity has been varied from 1.8 to 2.9 L feedstock/L catalyst/hr depending on the feedstock. The product fuel gas contains 40% to 55% methane, 35% to 50% carbon dioxide, and 5% to 10% hydrogen with as much as 2% ethane, but less than 0.1% ethylene or carbon monoxide, and small amounts of higher hydrocarbons. The byproduct water stream carries residual organics amounting to less than 500 mg/L COD. 9 refs., 1 fig., 4 tabs.

  2. Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-01-26

    In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.

  3. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

    2016-08-01

    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

  4. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition as part of a fuel meat thickness optimization effort for reactor performance other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  5. A review on the utilization of the Japan materials testing reactor (JMTR)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Kang, Y. H.; Kim, B. G.; Choo, K. N.; Oh, J. M.; Park, S. J.; Shin, Y. T

    1999-04-01

    The HANARO has possessed the potential capability for the testing of materials and fuels since the beginning of its operation in 1995. Recently, this reactor has contributed to various activities in nuclear power research in Korea. We need the recent technical data of developed countries to support these activities in nuclear power. Most of the developed countries in nuclear power have more than thirty years' experience in the irradiation test of nuclear fuel and material for performing their complicated in-core measurements of the change of material properties. They also have developed various types of sensors, equipment and techniques. This report describes the status of utilization of the irradiation facilities of the Japan Materials Testing Reactor(JMTR). It also describes the recent efforts of the JMTR in order to develop new irradiation test techniques. It will be our great pleasure for this report to help a broad range of people understand the generic contents (JMTR utilization, new techniques) of the JMTR. (author)

  6. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  7. Status of the Combined Third and Fourth NGNP Fuel Irradiations In the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti; Michael E. Davenport

    2013-07-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in September 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. Since the purpose of this combined experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  8. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-[TEMPERATURE GAS-COOLED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    A point design for a graphite-moderated, high-temperature, gas-cooled test reactor (HTG TR) has been developed by Idaho National Laboratory (INL) as part of a United States (U.S.) Department of Energy (DOE) initiative to explore and potentially expand the existing U.S. test reactor capabilities. This paper provides a summary of the design and its main attributes. The 200 MW HTG TR is a thermal-neutron spectrum reactor composed of hexagonal prismatic fuel and graphite reflector blocks. Twelve fuel columns (96 fuel blocks total and 6.34 m active core height) are arranged in two hexagonal rings to form a relatively compact, high-power density, annular core sandwiched between inner, outer, top, and bottom graphite reflectors. The HTG-TR is designed to operate at 7 MPa with a coolant inlet/outlet temperature of 325°C/650°C, and utilizes TRISO particle fuel from the DOE AGR Program with 425 ?m uranium oxycarbide (UCO) kernels and an enrichment of 15.5 wt% 235U. The primary mission of the HTG TR is material irradiation and therefore the core has been specifically designed and optimized to provide the highest possible thermal and fast neutron fluxes. The highest thermal neutron flux (3.90E+14 n/cm2s) occurs in the outer reflector, and the maximum fast flux levels (1.17E+14 n/cm2s) are produced in the central reflector column where most of the graphite has been removed. Due to high core temperatures under accident conditions, all the irradiation test facilities have been located in the inner and outer reflectors where fast flux levels decline. The core features a large number of irradiation positions with large test volumes and long test lengths, ideal for thermal neutron irradiation of large test articles. The total available test volume is more than 1100 liters. Up to four test loop facilities can be accommodated with pressure tube boundaries to isolate test articles and test fluids (e.g., liquid metal, liquid salt, light water) from the helium primary coolant system.

  9. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation.

  10. Application of the small punch test to reactor pressure vessel integrity

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [EPRI Nondestructive Evaluation Center, Charlotte, NC (United States); Viswanathan, R. [EPRI, Palo Alto, CA (United States); Foulds, J.R. [Failure Analysis Associates, Inc., Menlo Park, CA (United States)

    1998-07-01

    Based on prior success with fossil plant steels, EPRI is investigating the feasibility of applying the Small Punch test to determine the fracture toughness (K{sub ic}) of irradiated reactor pressure vessel (RPV) materials. The small punch test specimen is sufficiently small to alleviate future surveillance material availability concerns, as well as provide a means of direct vessel material interrogation by non-disruptive miniature sample removal and testing. A limited series of small punch tests on unirradiated and irradiated RPV steel materials has shown that the method can be used to estimate ductile-to-brittle transition temperatures and to determine the material fracture toughness (K{sub lc}, J{sub lc}). The results to date are described and the experimental difficulties that need to be resolved in achieving valid results are identified. (authors)

  11. Design of the VISTA-ITL Test Facility for an Integral Type Reactor of SMART and a Post-Test Simulation of a SBLOCA Test

    OpenAIRE

    2014-01-01

    To validate the performance and safety of an integral type reactor of SMART, a thermal-hydraulic integral effect test facility, VISTA-ITL, is introduced with a discussion of its scientific design characteristics. The VISTA-ITL was used extensively to assess the safety and performance of the SMART design, especially for its passive safety system such as a passive residual heat removal system, and to validate various thermal-hydraulic analysis codes. The VISTA-ITL program includes several tests...

  12. First Test of Lorentz Violation with a Reactor-based Antineutrino Experiment

    CERN Document Server

    Abe, Y; Anjos, J C dos; Bergevin, M; Bernstein, A; Bezerra, T J C; Bezrukhov, L; Blucher, E; Bowden, N S; Buck, C; Busenitz, J; Cabrera, A; Caden, E; Camilleri, L; Carr, R; Cerrada, M; Chang, P -J; Chimenti, P; Classen, T; Collin, A P; Conover, E; Conrad, J M; Crespo-Anadón, J I; Crum, K; Cucoanes, A; D'Agostino, M V; Damon, E; Dawson, J V; Dazeley, S; Dietrich, D; Djurcic, Z; Dracos, M; Durand, V; Ebert, J; Efremenko, Y; Elnimr, M; Erickson, A; Fallot, M; Fechner, M; von Feilitzsch, F; Felde, J; Fischer, V; Franco, D; Franke, A J; Franke, M; Furuta, H; Gama, R; Gil-Botella, I; Giot, L; Göger-Neff, M; Gonzalez, L F G; Goodman, M C; Goon, J TM; Greiner, D; Haag, N; Habib, S; Hagner, C; Hara, T; Hartmann, F X; Haser, J; Hatzikoutelis, A; Hayakawa, T; Hofmann, M; Horton-Smith, G A; Ishitsuka, M; Jochum, J; Jollet, C; Jones, C L; Kaether, F; Kalousis, L N; Kamyshkov, Y; Kaplan, D M; Katori, T; Kawasaki, T; Keefer, G; Kemp, E; de Kerret, H; Konno, T; Kryn, D; Kuze, M; Lachenmaier, T; Lane, C E; Lasserre, T; Letourneau, A; Lhuillier, D; Lima, H P; Lindner, M; López-Castanõ, J M; LoSecco, J M; Lubsandorzhiev, B K; Lucht, S; McKee, D; Maeda, J; Maesano, C N; Mariani, C; Maricic, J; Martino, J; Matsubara, T; Mention, G; Meregaglia, A; Meyer, M; Miletic, T; Milincic, R; Miyata, H; Mueller, Th A; Nagasaka, Y; Nakajima, K; Novella, P; Obolensky, M; Oberauer, L; Onillon, A; Osborn, A; Ostrovskiy, I; Palomares, C; Pepe, I M; Perasso, S; Perrin, P; Pfahler, P; Porta, A; Potzel, W; Pronost, G; Reichenbacher, J; Reinhold, B; Remoto, A; Röhling, M; Roncin, R; Roth, S; Rybolt, B; Sakamoto, Y; Santorelli, R; Sato, F; Schönert, S; Schoppmann, S; Schwetz, T; Shaevitz, M H; Shrestha, D; Sida, J -L; Sinev, V; Skorokhvatov, M; Smith, E; Spitz, J; Stahl, A; Stancu, I; Stokes, L F F; Strait, M; Stüken, A; Suekane, F; Sukhotin, S; Sumiyoshi, T; Sun, Y; Terao, K; Tonazzo, A; Toups, M; Thi, H H Trinh; Valdiviesso, G; Veyssiere, C; Wagner, S; Watanabe, H; White, B; Wiebusch, C; Winslow, L; Worcester, M; Wurm, M; Yanovitch, E; Yermia, F; Zimmer, V

    2012-01-01

    We present a search for Lorentz violation with 8249 candidate electron antineutrino events taken by the Double Chooz experiment in 227.9 live days of running. This analysis, featuring a search for a sidereal time dependence of the events, is the first test of Lorentz invariance using a reactor-based antineutrino source. No sidereal variation is present in the data and the disappearance results are consistent with sidereal time independent oscillations. Under the Standard-Model Extension (SME), we set the first limits on fourteen Lorentz violating coefficients associated with transitions between electron and tau flavor, and set two competitive limits associated with transitions between electron and muon flavor.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-03-01

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

  14. Evaluation of dynamic fracture toughness for Yong Gwang unit 5 reactor pressure vessel materials (Baseline Tests)

    Energy Technology Data Exchange (ETDEWEB)

    Chi Se Hwan; Kim, Joo Hag; Hong, Jun Hwa; Kwon, Sun Chil; Lee, Bong Sang [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    The dynamic fracture toughness (K{sub d}) of intermediate shell and its weld in SA 508 CI. 3 Yong Gwang 5 reactor pressure vessel was determined and evaluated. Precracked thirty six Charpy specimens were tested by using an instrumented impact tester. The purpose of present work is to evaluate and confirm the un-irradiated dynamic fracture toughness and to provide pre-irradiation baseline data for future evaluation on dynamic fracture toughness change during operation. 18 refs., 5 figs., 5 tabs. (Author)

  15. Evaluation of zinc addition on fuel cladding corrosion at the Halden test reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Kolstad, E.; Symons, W.J.; Bryhn-Integrigtsen, K.; Oberlaender, B.C.

    1996-08-01

    Experimental studies have shown that addition of zinc to a PWR environment reduces the general corrosion rates of materials in the primary system and delays the initiation of primary water stress corrosion cracking (PWSCC) in Alloy 600. In order to provide an early warning of any potential adverse effects on the fuel cladding, corrosion studies were initiated at the Halden test reactor. These tests were carried out in a PWR rig inserted in the Halden reactor core. The rig simulated thermal hydraulic and coolant conditions typical of a MR. It had two flow channels where the fuel rod segments were exposed to the coolant under irradiation flux. Selected pre-characterized rodlets with fresh and pre-irradiated standard and low-tin Zircaloy-4 material were irradiated for three cycles. First cycle lasted for 110 effective full power days (EFPDs), the second for 95 EFPDs and the last 62 EFPDs. The cladding corrosion behavior was monitored by initial, interim and final oxide thickness measurements by eddy current lift-off probe. Crud sampling was performed in both channels after cycle 1 and 2. Destructive post-irradiation examinations (PIE) of two rodlets, irradiated during cycle 1 and 2, have also been completed at the conclusion of the in-pile testing. This report presents the results on oxide thickness measurements, irradiation history and water chemistry data, and the PIE.

  16. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    Energy Technology Data Exchange (ETDEWEB)

    MH Lane

    2006-02-15

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

  17. Completion summary for borehole USGS 136 near the Advanced Test Reactor Complex, Idaho National Laboratory, Idaho

    Science.gov (United States)

    Twining, Brian V.; Bartholomay, Roy C.; Hodges, Mary K.V.

    2012-01-01

    In 2011, the U.S. Geological Survey, in cooperation with the U.S. Department of Energy, cored and completed borehole USGS 136 for stratigraphic framework analyses and long-term groundwater monitoring of the eastern Snake River Plain aquifer at the Idaho National Laboratory. The borehole was initially cored to a depth of 1,048 feet (ft) below land surface (BLS) to collect core, open-borehole water samples, and geophysical data. After these data were collected, borehole USGS 136 was cemented and backfilled between 560 and 1,048 ft BLS. The final construction of borehole USGS 136 required that the borehole be reamed to allow for installation of 6-inch (in.) diameter carbon-steel casing and 5-in. diameter stainless-steel screen; the screened monitoring interval was completed between 500 and 551 ft BLS. A dedicated pump and water-level access line were placed to allow for aquifer testing, for collecting periodic water samples, and for measuring water levels. Geophysical and borehole video logs were collected after coring and after the completion of the monitor well. Geophysical logs were examined in conjunction with the borehole core to describe borehole lithology and to identify primary flow paths for groundwater, which occur in intervals of fractured and vesicular basalt. A single-well aquifer test was used to define hydraulic characteristics for borehole USGS 136 in the eastern Snake River Plain aquifer. Specific-capacity, transmissivity, and hydraulic conductivity from the aquifer test were at least 975 gallons per minute per foot, 1.4 × 105 feet squared per day (ft2/d), and 254 feet per day, respectively. The amount of measureable drawdown during the aquifer test was about 0.02 ft. The transmissivity for borehole USGS 136 was in the range of values determined from previous aquifer tests conducted in other wells near the Advanced Test Reactor Complex: 9.5 × 103 to 1.9 × 105 ft2/d. Water samples were analyzed for cations, anions, metals, nutrients, total organic

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

    Energy Technology Data Exchange (ETDEWEB)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

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

  19. Supported Pd-Au Membrane Reactor for Hydrogen Production: Membrane Preparation, Characterization and Testing.

    Science.gov (United States)

    Iulianelli, Adolfo; Alavi, Marjan; Bagnato, Giuseppe; Liguori, Simona; Wilcox, Jennifer; Rahimpour, Mohammad Reza; Eslamlouyan, Reza; Anzelmo, Bryce; Basile, Angelo

    2016-05-09

    A supported Pd-Au (Au 7wt%) membrane was produced by electroless plating deposition. Permeation tests were performed with pure gas (H₂, H₂, N₂, CO₂, CH₄) for long time operation. After around 400 h under testing, the composite Pd-Au membrane achieved steady state condition, with an H₂/N₂ ideal selectivity of around 500 at 420 °C and 50 kPa as transmembrane pressure, remaining stable up to 1100 h under operation. Afterwards, the membrane was allocated in a membrane reactor module for methane steam reforming reaction tests. As a preliminary application, at 420 °C, 300 kPa of reaction pressure, space velocity of 4100 h(-1), 40% methane conversion and 35% hydrogen recovery were reached using a commercial Ni/Al₂O₃ catalyst. Unfortunately, a severe coke deposition affected irreversibly the composite membrane, determining the loss of the hydrogen permeation characteristics of the supported Pd-Au membrane.

  20. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; R. Schley; J. Palmer; K. Condie

    2014-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.

  1. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; J. E. Daw

    2011-03-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

  2. Non-Nuclear Testing of Compact Reactor Technologies at NASA MSFC

    Science.gov (United States)

    Houts, Michael G.; Pearson, J. Boise; Godfroy, Thomas J.

    2011-01-01

    Safe, reliable, compact, autonomous, long-life fission systems have numerous potential applications, both terrestrially and in space. Technologies and facilities developed in support of these systems could be useful to a variety of concepts. At moderate power levels, fission systems can be designed to operate for decades without the need for refueling. In addition, fast neutron damage to cladding and structural materials can be maintained at an acceptable level. Nuclear design codes have advanced to the stage where high confidence in the behavior and performance of a system can be achieved prior to initial testing. To help ensure reactor affordability, an optimal strategy must be devised for development and qualification. That strategy typically involves a combination of non-nuclear and nuclear testing. Non-nuclear testing is particularly useful for concepts in which nuclear operating characteristics are well understood and nuclear effects such as burnup and radiation damage are not likely to be significant. To be mass efficient, a SFPS must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial reactors. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference SFPS uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance. In addition, technologies from the SFPS system could be applicable to compact terrestrial systems. Recent non-nuclear testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference SFPS and evaluate methods for system integration. In July, 2011 an Annular Linear Induction Pump (ALIP) provided by Idaho National Laboratory was tested at the EFF-TF to assess performance and verify suitability for use in a10 kWe technology

  3. Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters

    Energy Technology Data Exchange (ETDEWEB)

    Darmann, Frank [Zenergy Power, Inc., Burlingame, CA (United States); Lombaerde, Robert [Zenergy Power, Inc., Burlingame, CA (United States); Moriconi, Franco [Zenergy Power, Inc., Burlingame, CA (United States); Nelson, Albert [Zenergy Power, Inc., Burlingame, CA (United States)

    2012-03-01

    Zenergy Power has successfully designed, built, tested, and installed in the US electrical grid a saturable reactor Fault Current Limiter. Beginning in 2007, first as SC Power Systems and from 2008 as Zenergy Power, Inc., ZP used DOE matching grant and ARRA funds to help refine the design of the saturated reactor fault current limiter. ZP ultimately perfected the design of the saturated reactor FCL to the point that ZP could reliably design a suitable FCL for most utility applications. Beginning with a very basic FCL design using 1G HTS for a coil housed in a LN2 cryostat for the DC bias magnet, the technology progressed to a commercial system that was offered for sale internationally. Substantial progress was made in two areas. First, the cryogenics cooling system progressed from a sub-cooled liquid nitrogen container housing the HTS coils to cryostats utilizing dry conduction cooling and reaching temperatures down to less than 20 degrees K. Large, round cryostats with warm bore diameters of 1.7 meters enabled the design of large tanks to hold the AC components. Second, the design of the AC part of the FCL was refined from a six legged spider design to a more compact and lighter design with better fault current limiting capability. Further refinement of the flux path and core shape led to an efficient saturated reactor design requiring less Ampere-turns to saturate the core. In conclusion, the development of the saturable reactor FCL led to a more efficient design not requiring HTS magnets and their associated peripheral equipment, which yielded a more economical product in line with the electric utility industry expectations. The original goal for the DOE funding of the ZP project Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters was to stimulate the HTS wire industry with, first 1G, then 2G, HTS wire applications. Over the approximately 5 years of ZP's product development program, the amount of HTS

  6. Application of Simulated Reactivity Feedback in Nonnuclear Testing of a Direct-Drive Gas-Cooled Reactor

    Science.gov (United States)

    Bragg-Sitton, S. M.; Webster, K. L.

    2007-01-01

    Nonnuclear testing can be a valuable tool in the development of an in-space nuclear power or propulsion system. In a nonnuclear test facility, electric heaters are used to simulate heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response and response characteristics, and assess potential design improvements with a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE 100a heat pipe cooled, electrically heated reactor and heat exchanger hardware. This Technical Memorandum discusses the status of the planned dynamic test methodology for implementation in the direct-drive gas-cooled reactor testing and assesses the additional instrumentation needed to implement high-fidelity dynamic testing.

  7. Premiere toob lavale jalgpallimeeskonna, inimkatsed ja punase tooli / Kairi Prints

    Index Scriptorium Estoniae

    Prints, Kairi, 1977-

    2012-01-01

    Premiere 2012 osalevad neli Eesti tantsukunstnikku: Svetlana Grigorjeva tantsulavastusega "sõp rus est", Kaisa Selde, Kristina-Maria Heinsalu ja Christin Lunts tantsulavastusega "fie", esmakordselt võtab osa välismaalane - sakslanna Mareike Franz tantsulavastusega "Duett". Kõik esietenduvad 9. veebruaril Kanuti gildi saalis

  8. Premiere toob lavale jalgpallimeeskonna, inimkatsed ja punase tooli / Kairi Prints

    Index Scriptorium Estoniae

    Prints, Kairi, 1977-

    2012-01-01

    Premiere 2012 osalevad neli Eesti tantsukunstnikku: Svetlana Grigorjeva tantsulavastusega "sõp rus est", Kaisa Selde, Kristina-Maria Heinsalu ja Christin Lunts tantsulavastusega "fie", esmakordselt võtab osa välismaalane - sakslanna Mareike Franz tantsulavastusega "Duett". Kõik esietenduvad 9. veebruaril Kanuti gildi saalis

  9. Open Access Publishing in Indian Premier Research Institutions

    Science.gov (United States)

    Bhat, Mohammad Hanief

    2009-01-01

    Introduction: Publishing research findings in open access journals is a means of enhancing visibility and consequently increasing the impact of publications. This study provides an overview of open access publishing in premier research institutes of India. Method: The publication output of each institution from 2003 to 2007 was ascertained through…

  10. Residual radioactivity guidelines for the heavy water components test reactor at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B. Smith, R.; McNeil, J.

    1997-04-01

    Guidelines were developed for acceptable levels of residual radioactivity in the Heavy Water Components Test Reactor (HWCTR) facility at the conclusion of its decommissioning. Using source terms developed from data generated in a detailed characterization study, the RESRAD and RASRAD-BUILD computer codes were used to calculate derived concentration guideline levels (DCGLs) for the radionuclides that will remain in the facility. The calculated DCGLs, when compared to existing concentrations of radionuclides measured during a 1996 characterization program, indicate that no decontamination of concrete surfaces will be necessary. Also, based on the results of the calculations, activated concrete in the reactor biological shield does not have to be removed, and imbedded radioactive piping in the facility can remain in place. Viewed in another way, the results of the calculations showed that the present inventory of residual radioactivity in the facility (not including that associated with the reactor vessel and steam generators) would produce less than one millirem per year above background to a hypothetical individual on the property. The residual radioactivity is estimated to be approximately 0.04 percent of the total inventory in the facility as of March, 1997. According to the results, the only radionuclides that would produce greater than 0.0.1-millirem per year are Am-241 (0.013 mrem/yr at 300 years), C-14 (0.022 mrem/yr at 1000 years) and U-238 (0.034 mrem/yr at 6000 years). Human exposure would occur only through the groundwater pathways, that is, from water drawn from, a well on the property. The maximum exposure would be approximately one percent of the 4 millirem per year ground water exposure limit established by the U.S. Environmental Protection Agency. 11 refs., 13 figs., 15 tabs.

  11. Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance, E706 (IIIC)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This test method describes the concept and use of helium accumulation for neutron fluence dosimetry for reactor vessel surveillance. Although this test method is directed toward applications in vessel surveillance, the concepts and techniques are equally applicable to the general field of neutron dosimetry. The various applications of this test method for reactor vessel surveillance are as follows: 1.1.1 Helium accumulation fluence monitor (HAFM) capsules, 1.1.2 Unencapsulated, or cadmium or gadolinium covered, radiometric monitors (RM) and HAFM wires for helium analysis, 1.1.3 Charpy test block samples for helium accumulation, and 1.1.4 Reactor vessel (RV) wall samples for helium accumulation. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  12. Neural network setpoint control of an advanced test reactor experiment loop simulation

    Energy Technology Data Exchange (ETDEWEB)

    Cordes, G.A.; Bryan, S.R.; Powell, R.H.; Chick, D.R.

    1990-09-01

    This report describes the design, implementation, and application of artificial neural networks to achieve temperature and flow rate control for a simulation of a typical experiment loop in the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory (INEL). The goal of the project was to research multivariate, nonlinear control using neural networks. A loop simulation code was adapted for the project and used to create a training set and test the neural network controller for comparison with the existing loop controllers. The results for three neural network designs are documented and compared with existing loop controller action. The neural network was shown to be as accurate at loop control as the classical controllers in the operating region represented by the training set. 9 refs., 28 figs., 2 tabs.

  13. Improving the AGR Fuel Testing Power Density Profile Versus Irradiation-Time in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gray S. Chang; David A. Petti; John T. Maki; Misti A. Lillo

    2009-05-01

    The Very High Temperature gas-cooled Reactor (VHTR), which is currently being developed, achieves simplification of safety through reliance on ceramic-coated fuel particles. Each TRISO-coated fuel particle has its own containment which serves as the principal barrier against radionuclide release under normal operating and accident conditions. These fuel particles, in the form of graphite fuel compacts, are currently undergoing a series of irradiation tests in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) to support the Advanced Gas-Cooled Reactor (AGR) fuel qualification program. A representive coated fuel particle with an 235U enrichment of 19.8 wt% was used in this analysis. The fuel burnup analysis tool used to perform the neutronics study reported herein, couples the Monte Carlo transport code MCNP, with the radioactive decay and burnup code ORIGEN2. The fuel burnup methodology known as Monte-Carlo with ORIGEN2 (MCWO) was used to evaluate the AGR experiment assembly and demonstrate compliance with ATR safety requirements. For the AGR graphite fuel compacts, the MCWO-calculated fission power density (FPD) due to neutron fission in 235U is an important design parameter. One of the more important AGR fuel testing requirements is to maintain the peak fuel compact temperature close to 1250°C throughout the proposed irradiation campaign of 550 effective full power days (EFPDs). Based on the MCWO-calculated FPD, a fixed gas gap size was designed to allow regulation of the fuel compact temperatures throughout the entire fuel irradiation campaign by filling the gap with a mixture of helium and neon gases. The chosen fixed gas gap can only regulate the peak fuel compact temperature in the desired range during the irradiation test if the ratio of the peak power density to the time-dependent low power density (P/T) at 550 EFPDs is less than 2.5. However, given the near constant neutron flux within the ATR driver core and the depletion of 235U in

  14. Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

    Science.gov (United States)

    Springborn, R. H.

    1971-01-01

    Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

  15. Verifying seismic design of nuclear reactors by testing. Volume 1: test plan

    Energy Technology Data Exchange (ETDEWEB)

    1979-07-20

    This document sets forth recommendations for a verification program to test the ability of operational nuclear power plants to achieve safe shutdown immediately following a safe-shutdown earthquake. The purpose of the study is to develop a program plan to provide assurance by physical demonstration that nuclear power plants are earthquake resistant and to allow nuclear power plant operators to (1) decide whether tests should be conducted on their facilities, (2) specify the tests that should be performed, and (3) estimate the cost of the effort to complete the recommended test program.

  16. Monte Carlo Simulation Study of a Differential Calorimeter Measuring the Nuclear Heating in Material Testing Reactors

    Science.gov (United States)

    Amharrak, H.; Reynard-Carette, C.; Lyoussi, A.; Carette, M.; Brun, J.; De Vita, C.; Fourmentel, D.; Villard, J.-F.; Guimbal, P.

    2016-02-01

    The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. Then these measurements are used for other materials, other geometries, or other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. This paper will present new simulations with MCNP Monte-Carlo transport code to determine the gamma heating profile inside the calorimeter. The whole complex geometry of the sensor has been considered. We use as an input source in the model, the photon spectra calculated in various positions of CARMEN-1 irradiation program in OSIRIS reactor. After a description of the differential calorimeter device, the MCNP modeling used for the calculations of radial profile of nuclear heating inside the calorimeter elements will be introduced. The obtained results of different simulations will be detailed and discussed in this paper. The charged particle equilibrium inside the calorimeter elements will be studied. Then we will focus on parametric studies of the various components of the calorimeter. The influence of source type will be also took into account. Moreover the influence of the material used for the sample will be described.

  17. Monte Carlo Simulation Study of a Differential Calorimeter Measuring the Nuclear Heating in Material Testing Reactors

    Directory of Open Access Journals (Sweden)

    Amharrak H.

    2016-01-01

    Full Text Available The nuclear heating measurements in Material Testing Reactors (MTRs are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. Then these measurements are used for other materials, other geometries, or other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. This paper will present new simulations with MCNP Monte-Carlo transport code to determine the gamma heating profile inside the calorimeter. The whole complex geometry of the sensor has been considered. We use as an input source in the model, the photon spectra calculated in various positions of CARMEN-1 irradiation program in OSIRIS reactor. After a description of the differential calorimeter device, the MCNP modeling used for the calculations of radial profile of nuclear heating inside the calorimeter elements will be introduced. The obtained results of different simulations will be detailed and discussed in this paper. The charged particle equilibrium inside the calorimeter elements will be studied. Then we will focus on parametric studies of the various components of the calorimeter. The influence of source type will be also took into account. Moreover the influence of the material used for the sample will be described.

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

    CERN Document Server

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

    2015-01-01

    A prototype of neutron spectrometer based on diamond detectors has been developed. This prototype consists of a $^6$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 $10^8$ n/cm$^2$s and at the 3 MeV D-D monochromatic neutron source na...

  19. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010

    Energy Technology Data Exchange (ETDEWEB)

    Rahmat Aryaeinejad; Douglas S. Crawford; Mark D. DeHart; George W. Griffith; D. Scott Lucas; Joseph W. Nielsen; David W. Nigg; James R. Parry; Jorge Navarro

    2010-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or “Core Modeling Update”) Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  20. Measurement of photon flux with a miniature gas ionization chamber in a Material Testing Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fourmentel, D., E-mail: damien.fourmentel@cea.fr [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Filliatre, P.; Villard, J.F.; Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Reynard-Carette, C. [Aix-Marseille Université, LISA EA 4672, cedex 20, Marseille 13397 (France); Carcreff, H. [CEA, DEN, DRSN, Saclay, F-91191 Gif-sur-Yvette (France)

    2013-10-01

    Nuclear heating measurements in Material Testing Reactors (MTR) are crucial for the design of the experimental devices and the prediction of the temperature of the hosted samples. Nuclear heating in MTR materials (except fuel) is mainly due to the energy deposition by the photon flux. Therefore, the photon flux is a key input parameter for the computer codes which simulate nuclear heating and temperature reached by samples/devices under irradiation. In the Jules Horowitz MTR under construction at the CEA Cadarache, the maximal expected nuclear heating levels will be about 15 to 18 W g{sup −1} and it will be necessary to assess this parameter with the best accuracy. An experiment was performed at the OSIRIS reactor to combine neutron flux, photon flux and nuclear heating measurements to improve the knowledge of the nuclear heating in MTR. There are few appropriate sensors for selective measurement of the photon flux in MTR even if studies and developments are ongoing. An experiment, called CARMEN-1, was conducted at the OSIRIS MTR and we used in particular a gas ionization chamber based on miniature fission chamber design to measure the photon flux. In this paper, we detail Monte-Carlo simulations to analyze the photon fluxes with ionization chamber measurements and we compare the photon flux calculations to the nuclear heating measurements. These results show a good accordance between photon flux measurements and nuclear heating measurement and allow improving the knowledge of these parameters.

  1. Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pope, M. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, M. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Morrell, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jamison, R. K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nef, E. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nigg, D. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses, a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.

  2. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2013

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg

    2013-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  3. Preliminary Feasibility, Design, and Hazard Analysis of a Boiling Water Test Loop Within the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Douglas M. Gerstner

    2009-05-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The ATR and its support facilities are located at the Idaho National Laboratory (INL). A Boiling Water Test Loop (BWTL) is being designed for one of the irradiation test positions within the. The objective of the new loop will be to simulate boiling water reactor (BWR) conditions to support clad corrosion and related reactor material testing. Further it will accommodate power ramping tests of candidate high burn-up fuels and fuel pins/rods for the commercial BWR utilities. The BWTL will be much like the pressurized water loops already in service in 5 of the 9 “flux traps” (region of enhanced neutron flux) in the ATR. The loop coolant will be isolated from the primary coolant system so that the loop’s temperature, pressure, flow rate, and water chemistry can be independently controlled. This paper presents the proposed general design of the in-core and auxiliary BWTL systems; the preliminary results of the neutronics and thermal hydraulics analyses; and the preliminary hazard analysis for safe normal and transient BWTL and ATR operation.

  4. Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

    Energy Technology Data Exchange (ETDEWEB)

    Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-11-30

    The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.

  5. A miniaturized test method for the mechanical characterization of structural materials for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gondi, P. [Rome-2 Univ. (Italy). Mech. Eng. Dept.; Donato, A. [ENEA CRE, Fusion Sector, Frascati, Rome (Italy); Montanari, R. [Rome-2 Univ. (Italy). Mech. Eng. Dept.; Sili, A. [Rome-2 Univ. (Italy). Mech. Eng. Dept.

    1996-10-01

    This work deals with a non-destructive method for mechanical tests which is based on the indentation of materials at a constant rate by means of a cylinder with a small radius and penetrating flat surface. The load versus penetration depth curves obtained using this method have shown correspondences with those of tensile tests and have given indications about the mechanical properties on a reduced scale. In this work penetration tests have been carried out on various kinds of Cr martensitic steels (MANET-2, BATMAN and modified F82H) which are of interest for first wall and structural applications in future fusion reactors. The load versus penetration depth curves have been examined with reference to data obtained in tensile tests and to microhardness measurements. Penetration tests have been performed at various temperature (from -180 to 100 C). Conclusions, which can be drawn for the ductile to brittle transition, are discussed for MANET-2 steel. Preliminary results obtained on BATMAN and modified F82H steels are reported. The characteristics of the indenter imprints have been studied by scanning electron microscopy. (orig.).

  6. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.

  7. Design of high temperature irradiation materials inspection cells. (Spent fuel inspection cells) in the High Temperature Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ino, Hiroichi; Ueta, Shouhei; Suzuki, Hiroshi; Sawa, Kazuhiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tobita, Tsutomu [Nuclear Engineering Company, Ltd., Tokai, Ibaraki (Japan)

    2002-01-01

    This report summarizes design requirements and design results for shields, ventilation system and fuel handling devices for the high temperature irradiation materials inspection cells (spent fuel inspection cells). These cells are small cells to carry out few post-irradiation examinations of spent fuels, specimen, etc., which are irradiated in the High Temperature Engineering Test Reactor, since the cells should be built in limited space in the HTTR reactor building, the cells are designed considering relationship between the cells and the reactor building to utilize the limited space effectively. The cells consist of three partitioned hot cells with wall for neutron and gamma-ray shields, ventilation system including filtering units and fuel handling devices. The post-irradiation examinations of the fuels and materials are planed by using the cells and the Hot Laboratory of the Japan Materials Testing Reactor to establish the technology basis on high temperature gas-cooled reactors (HTGRs). In future, irradiation tests and post-irradiation examinations will be carried out with the cells to upgrade present HTGR technologies and to make the innovative basic research on high-temperature engineering. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-02-15

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

  9. Model for collisional fast ion diffusion into Tokamak Fusion Test Reactor loss cone

    Energy Technology Data Exchange (ETDEWEB)

    Chang, C.S. [New York Univ., NY (United States). Courant Inst. of Mathematical Sciences]|[Korea Advanced Inst. of Science and Technology, Seoul (Korea, Republic of); Zweben, S.J.; Schivell, J.; Budny, R.; Scott, S. [Princeton Univ., NJ (United States). Plasma Physics Lab.

    1994-08-01

    An analytic model is developed to estimate the classical pitch angle scattering loss of energetic fusion product ions into prompt loss orbits in a tokamak geometry. The result is applied to alpha particles produced by deutrium-tritium fusion reactions in a plasma condition relevant to Tokamak Fusion Test Reactor (TFTR). A poloidal angular distribution of collisional fast ion loss at the first wall is obtained and the numerical result from the TRANSP code is discussed. The present model includes the effect that the prompt loss boundary moves away from the slowing-down path due to reduction in banana thickness, which enables us to understand, for the first time. the dependence of the collisional loss rate on Z{sub eff}.

  10. Safety Evaluation for Packaging for onsite Transfer of plutonium recycle test reactor ion exchange columns

    Energy Technology Data Exchange (ETDEWEB)

    Smith, R.J.

    1995-09-11

    The purpose of this Safety Evaluation for Packaging (SEP) is to authorize the use of three U.S. Department of Transportation (DOT) 7A, Type A metal boxes (Capital Industries Part No. S 0600-0600-1080- 0104) to package 12 Plutonium Recycle Test Reactor (PRTR) ion exchange columns as low-level waste (LLW). The packages will be transferred from the 309 Building in the 300 Area to low level waste burial in the 200 West Area. Revision 1 of WHC-SD-TP-SEP-035 (per ECN No. 621467) documents that the boxes containing ion exchange columns and grout will maintain the payload under normal conditions of transport if transferred without the box lids

  11. Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

    Energy Technology Data Exchange (ETDEWEB)

    Romano, T.

    1997-09-29

    This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

  12. SPECIFICATIONS FOR HIGH TEMPERATURE LATTICE TEST REACTOR BUILDING 318 PROJECT CAH-100

    Energy Technology Data Exchange (ETDEWEB)

    Vitro Engineering Company

    1964-07-15

    This is the specifications for the High Temperature Lattice Test Reactor Building 318 and it is divided into the following 21 divisions or chapters: (1) Excavating, Backfill & Grading; (2) Reinforced Concrete; (3) Masonry; (4) Structural Steel & Miscellaneous Metal Items, Contents - Division V; (5) Plumbing, Process & Service Piping; (6) Welding; (7) Insulated Metal Siding; (8) Roof Decks & Roofing; (9) Plaster Partitions & Ceiling; (10) Standard Doors, Windows & Hardware; (11) Shielding Doors; (12) Sprinkler System & Fire Extinguishers, Contents - Division XIII; (13) Heating, Ventilating & Air Conditioning; (14) Painting, Protective Coating & Floor Covering, Contents - Division XV; (15) Electrical; (16) Communications & Alarm Systems; (17) Special Equipment & Furnishings; (18) Overhead Bridge Crane; (19) Prefabricated Steel Building; (20) Paved Drive; and (21) Landscaping & Irrigation Sprinklers.

  13. Crack growth tests on a ferritic reactor pressure vessel steel under the simultaneous influence of simulated BWR coolant and irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, H. [VGB PowerTech e.V., Essen (Germany); Huettner, F. [Hamburgische Electricitaets-Werke AG, Hamburg (Germany); Ilg, U. [EnBW Kraftwerke AG, Philippsburg (Germany); Wachter, O. [E.ON Kernkraft GmbH, Hannover(Germany); Widera, M. [RWE Power AG, Essen (Germany); Brozova, A.; Ernestova, M.; Kysela, J.; Vsolak, R. [Nuclear Research Institute Rez plc (Czech Republic)

    2004-07-01

    Crack growth tests under constant load with initial in-situ cycling were performed on the low alloy reactor pressure vessel (RPV) steel 22 NiMoCr 3 7 (A 508 Cl. 2) with the goal to determine crack growth rates of irradiated and non-irradiated steel under the simultaneous influence of simulated BWR coolant and irradiation. The tests were performed under conditions as near as possible to operational conditions in a commercial BWR reactor. The research results are summarized and are compared with international data. (orig.)

  14. UO2 and PuO2 utilization in high temperature engineering test reactor with helium coolant

    Science.gov (United States)

    Waris, Abdul; Aji, Indarta K.; Novitrian, Pramuditya, Syeilendra; Su'ud, Zaki

    2016-03-01

    High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO2 fuel. In this study, we have evaluated the use of UO2 and PuO2 in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of 235U in loaded fuel is 18.0% or above.

  15. Summary of experimental tests of elastomeric seismic isolation bearings for use in nuclear reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Seidensticker, R.W.; Chang, Y.W.; Kulak, R.F.

    1992-05-01

    This paper describes an experimental test program for isolator bearings which was developed to help establish the viability of using laminated elastomer bearings for base isolation of nuclear reactor plants. The goal of the test program is to determine the performance characteristics of laminated seismic isolation bearings under a wide range of loadings. Tests were performed on scale-size laminated seismic isolators both within the design shear strain range to determine the response of the bearing under expected earthquake loading conditions, and beyond the design range to determine failure modes and to establish safety margins. Three types of bearings, each produced from a different manufacturer, have been tested: (1) high shape factor-high damping-high shear modulus bearings; (2) medium shape factor-high damping-high shear modulus bearings; and (3) medium shape factor-high damping-low shear modulus bearings. All of these tests described in this report were performed at the Earthquake Engineering Research Center at the University of California, Berkeley, with technical assistance from ANL. The tests performed on the three types of bearings have confirmed the high performance characteristics of the high damping-high and low shear modulus elastomeric bearings. The bearings have shown that they are capable of having extremely large shear strains before failure occurs. The most common failure mechanism was the debonding of the top steel plate from the isolators. This failure mechanism can be virtually eliminated by improved manufacturing quality control. The most important result of the failure test of the isolators is the fact that bearings can sustain large horizontal displacement, several times larger than the design value, with failure. Their performance in moderate and strong earthquakes will be far superior to conventional structures.

  16. Summary of experimental tests of elastomeric seismic isolation bearings for use in nuclear reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Seidensticker, R.W.; Chang, Y.W.; Kulak, R.F.

    1992-01-01

    This paper describes an experimental test program for isolator bearings which was developed to help establish the viability of using laminated elastomer bearings for base isolation of nuclear reactor plants. The goal of the test program is to determine the performance characteristics of laminated seismic isolation bearings under a wide range of loadings. Tests were performed on scale-size laminated seismic isolators both within the design shear strain range to determine the response of the bearing under expected earthquake loading conditions, and beyond the design range to determine failure modes and to establish safety margins. Three types of bearings, each produced from a different manufacturer, have been tested: (1) high shape factor-high damping-high shear modulus bearings; (2) medium shape factor-high damping-high shear modulus bearings; and (3) medium shape factor-high damping-low shear modulus bearings. All of these tests described in this report were performed at the Earthquake Engineering Research Center at the University of California, Berkeley, with technical assistance from ANL. The tests performed on the three types of bearings have confirmed the high performance characteristics of the high damping-high and low shear modulus elastomeric bearings. The bearings have shown that they are capable of having extremely large shear strains before failure occurs. The most common failure mechanism was the debonding of the top steel plate from the isolators. This failure mechanism can be virtually eliminated by improved manufacturing quality control. The most important result of the failure test of the isolators is the fact that bearings can sustain large horizontal displacement, several times larger than the design value, with failure. Their performance in moderate and strong earthquakes will be far superior to conventional structures.

  17. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for FHRs

    Science.gov (United States)

    Lu, Qiuping

    Direct Reactor Auxiliary Cooling System (DRACS) is a passive decay heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines coated particle fuel and a graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops, relying completely on buoyancy as the driving force. These loops are coupled through two heat exchangers, namely, the DRACS Heat Exchanger (DHX) and the Natural Draft Heat Exchanger (NDHX). In addition, a fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during normal operation of the reactor, but to keep the DRACS ready for activation, if needed, during accidents. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for FHRs built or tested in the literature. The primary goal of the present research is to design, test, and model the DRACS for FHR applications. Previously, a detailed modular design of the DRACS for a 20-MWth FHR was developed. As a starting point, the DRACS was designed to remove 1% of the reactor nominal power, i.e., 200 kW decay power. In addition, a detailed scaling analysis has been performed to develop the key non-dimensional numbers that characterize the DRACS system. Based on the previous work on the prototypic DRACS design and scaling analysis, two scaled-down test facilities have been designed and constructed, namely, Low-temperature DRACS Test Facility (LTDF) and High-temperature DRACS Test Facility (HTDF). The LTDF has a nominal power capacity of 6 kW. It uses 1.0-MPa water as the primary coolant, 0.1-MPa water as the secondary coolant, and ambient air as the ultimate heat sink. The main purpose of the LTDF is to examine the couplings among the three natural circulation/convection loops in the DRACS, as well as to provide design and operation experience for the HTDF. An extensive test matrix has

  18. In-situ Creep Testing Capability Development for Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    B. G. Kim; J. L. Rempe; D. L. Knudson; K. G. Condie; B. H. Sencer

    2010-08-01

    Creep is the slow, time-dependent strain that occurs in a material under a constant strees (or load) at high temperature. High temperature is a relative term, dependent on the materials being evaluated. A typical creep curve is shown in Figure 1-1. In a creep test, a constant load is applied to a tensile specimen maintained at a constant temperature. Strain is then measured over a period of time. The slope of the curve, identified in the figure below, is the strain rate of the test during Stage II or the creep rate of the material. Primary creep, Stage I, is a period of decreasing creep rate due to work hardening of the material. Primary creep is a period of primarily transient creep. During this period, deformation takes place and the resistance to creep increases until Stage II, Secondary creep. Stage II creep is a period with a roughly constant creep rate. Stage II is referred to as steady-state creep because a balance is achieved between the work hardening and annealing (thermal softening) processes. Tertiary creep, Stage III, occurs when there is a reduction in cross sectional area due to necking or effective reduction in area due to internal void formation; that is, the creep rate increases due to necking of the specimen and the associated increase in local stress.

  19. Premiere of Film Nine-Mile Fragrance Held in Kenya

    Institute of Scientific and Technical Information of China (English)

    2012-01-01

    <正>The United Nations Environment Program (UNEP) and the United Nations Human Settlement Program (UN-HABITAT) in Kenya staged the premiere of the film Nine-Mile Fragrance depicting the Wenchuan earthquake relief efforts. The film, showing the unique culture of the Qiang ethnic group and the great spirit of love of the Chinese people in their earthquake relief work, left a deep impression on the African audience.

  20. Supported Pd-Au Membrane Reactor for Hydrogen Production: Membrane Preparation, Characterization and Testing

    Directory of Open Access Journals (Sweden)

    Adolfo Iulianelli

    2016-05-01

    Full Text Available A supported Pd-Au (Au 7wt% membrane was produced by electroless plating deposition. Permeation tests were performed with pure gas (H2, H2, N2, CO2, CH4 for long time operation. After around 400 h under testing, the composite Pd-Au membrane achieved steady state condition, with an H2/N2 ideal selectivity of around 500 at 420 °C and 50 kPa as transmembrane pressure, remaining stable up to 1100 h under operation. Afterwards, the membrane was allocated in a membrane reactor module for methane steam reforming reaction tests. As a preliminary application, at 420 °C, 300 kPa of reaction pressure, space velocity of 4100 h−1, 40% methane conversion and 35% hydrogen recovery were reached using a commercial Ni/Al2O3 catalyst. Unfortunately, a severe coke deposition affected irreversibly the composite membrane, determining the loss of the hydrogen permeation characteristics of the supported Pd-Au membrane.

  1. Recirculation pump discharge line break tests at ROSA-III for a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, M.; Anoda, Y.; Kumamaru, H.; Nakamura, H.; Shiba, M.; Tasaka, K.

    1985-08-01

    Three loss-of-coolant accident (LOCA) tests were conducted at the Rig of Safety Assessment (ROSA)-III test facility, which simulates boiling water reactor (BWR)/6-251 with a volumetric scaling factor of 1/424. The fundamental features of the recirculation pump discharge line break LOCA and the effects of break areas on the features are investigated. It has been confirmed experimentally that the LOCA phenomena in the discharge line break are analogous to those in the suction line break with the same effective choking flow area, which is a sum of the least choking flow areas along the break flow paths and controls the system pressure responses. In general, the maximum effective choking flow area is (A /SUB j/ + A /SUB p/ ) for discharge line breaks and (A /SUB j/ + A /SUB o/ ) for suction line breaks, where A /SUB j/ , A /SUB p/ , and A /SUB o/ are the flow areas of the jet pump drive nozzles, the main recirculation pump discharge nozzle, and the break, respectively. The similarity between the ROSA-III test and a BWR LOCA has been confirmed in the key phenomena by the analyses using the RELAP5/MOD1 code. An atypical behavior is observed in the fuel rod surface temperature transient in the early phase of blowdown due to the limitation of the ROSA-III initial core power.

  2. Long Term Out-of-pile Thermocouple Tests in Conditions Representative for Nuclear Gas-cooled High Temperature Reactors

    OpenAIRE

    LAURIE Mathias; FOURREZ Stephane; FUETTERER Michael; LAPETITE Jean-Marc

    2011-01-01

    During irradiation tests at high temperature, failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. To understand and remediate this problem, out-ofpile tests were performed with thermocouples in carburizing atmospheres which can be assumed to be at least locally representative for High Temperature Reactors. The objective was to screen those thermocouples which would consecutively be used under irradiation. Two such screening tests have been performed with a set of...

  3. Fabrication and Comparison of Fuels for Advanced Gas Reactor Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey Phillips; Charles Barnes; John Hunn

    2010-10-01

    As part of the program to demonstrate TRISO-coated fuel for the Next Generation Nuclear Plant, a series of irradiation tests of Advanced Gas Reactor (AGR) fuel are being performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. In the first test, called “AGR-1,” graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 until November 2009. Development of AGR-1 fuel sought to replicate the properties of German TRISO-coated particles. No particle failures were seen in the nearly 3-year irradiation to a burn up of 19%. The AGR-1 particles were coated in a two-inch diameter coater. Following fabrication of AGR-1 fuel, process improvements and changes were made in each of the fabrication processes. Changes in the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a six-inch diameter coater using a change size about twenty-one times that of the two-inch diameter coater used to coat AGR-1 particles. Changes were also made in the compacting process, including increasing the temperature and pressure of pressing and using a different type of press. Irradiation of AGR-2 fuel began in late spring 2010. Properties of AGR-2 fuel compare favorably with AGR-1 and historic German fuel. Kernels are more homogeneous in shape, chemistry and density. TRISO-particle sphericity, layer thickness standard deviations, and defect fractions are also comparable. In a sample of 317,000 particles from deconsolidated AGR-2 compacts, 3 exposed kernels were found in a leach test. No SiC defects were found in a sample of 250,000 deconsolidated particles, and no IPyC defects in a sample of 64,000 particles. The primary difference in properties between AGR-1 and AGR-2 compacts is that AGR-2 compacts have a higher matrix density, 1.6 g/cm3 compared to about 1.3 g/cm3 for AGR-1 compacts. Based on

  4. Final report on in-reactor tensile tests on OFHC - Copper and CuCrZr alloy

    DEFF Research Database (Denmark)

    Singh, B.N; Edwards, D.J.; Tähtinen, S.

    2004-01-01

    of uniaxial tensile tests on pure copper and a CuCrZr alloy in a fission reactor at 363 and 393K. In the following, we first describe the experiments and then present results illustrating the build up ofstress as a function of concurrently increasing strain and displacement dose level. Results on both pre...

  5. THE COMPONENT TEST FACILITY – A NATIONAL USER FACILITY FOR TESTING OF HIGH TEMPERATURE GAS-COOLED REACTOR (HTGR) COMPONENTS AND SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    David S. Duncan; Vondell J. Balls; Stephanie L. Austad

    2008-09-01

    The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.

  6. Hydrothermal Testing of K Basin Sludge and N Reactor Fuel at Sludge Treatment Project Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, Calvin H.; Schmidt, Andrew J.; Thornton, Brenda M.

    2007-03-30

    The Sludge Treatment Project (STP), managed for the U. S. DOE by Fluor Hanford (FH), was created to design and operate a process to eliminate uranium metal from K Basin sludge prior to packaging for Waste Isolation Pilot Plant (WIPP). The STP process uses high temperature liquid water to accelerate the reaction, produce uranium dioxide from the uranium metal, and safely discharge the hydrogen. Under nominal process conditions, the sludge will be heated in pressurized water at 185°C for as long as 72 hours to assure the complete reaction (corrosion) of up to 0.25-inch diameter uranium metal pieces. Under contract to FH, the Pacific Northwest National Laboratory (PNNL) conducted bench-scale testing of the STP hydrothermal process in November and December 2006. Five tests (~50 ml each) were conducted in sealed, un-agitated reaction vessels under the hydrothermal conditions (e.g., 7 to 72 h at 185°C) of the STP corrosion process using radioactive sludge samples collected from the K East Basin and particles/coupons of N Reactor fuel also taken from the K Basins. The tests were designed to evaluate and understand the chemical changes that may be occurring and the effects that any changes would have on sludge rheological properties. The tests were not designed to evaluate engineering aspects of the process. The hydrothermal treatment affected the chemical and physical properties of the sludge. In each test, significant uranium compound phase changes were identified, resulting from dehydration and chemical reduction reactions. Physical properties of the sludge were significantly altered from their initial, as-settled sludge values, including, shear strength, settled density, weight percent water, and gas retention.

  7. Testing of a 7-tube palladium membrane reactor for potential use in TEP

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, Bryan J [Los Alamos National Laboratory; Trujillo, Stephen [Los Alamos National Laboratory; Willms, R. Scott [Los Alamos National Laboratory

    2010-01-01

    A Palladium Membrane Reactor (PMR) consists of a palladium/silver membrane permeator filled with catalyst (catalyst may be inside or outside the membrane tubes). The PMR is designed to recover tritium from the methane, water, and other impurities present in fusion reactor effluent. A key feature of a PMR is that the total hydrogen isotope content of a stream is significantly reduced as (1) methane-steam reforming and/or water-gas shift reactions proceed on the catalyst bed and (2) hydrogen isotopes are removed via permeation through the membrane. With a PMR design matched to processing requirements, nearly complete hydrogen isotope removals can be achieved. A 3-tube PMR study was recently completed. From the results presented in this study, it was possible to conclude that a PMR is appropriate for TEP, perforated metal tube protectors function well, platinum on aluminum (PtA) catalyst performs the best, conditioning with air is probably required to properly condition the Pd/Ag tubes, and that CO/CO{sub 2} ratios maybe an indicator of coking. The 3-tube PMR had a permeator membrane area of 0.0247 m{sup 2} and a catalyst volume to membrane area ratio of 4.63 cc/cm{sup 2} (with the catalyst on the outside of the membrane tubes and the catalyst only covering the membrane tube length). A PMR for TEP will require a larger membrane area (perhaps 0.35 m{sup 2}). With this in mind, an intermediate sized PMR was constructed. This PMR has 7 permeator tubes and a total membrane area of 0.0851 m{sup 2}. The catalyst volume to membrane area ratio for the 7-tube PMR was 5.18 cc/cm{sup 2}. The total membrane area of the 7-tube PMR (0.0851 m{sup 2}) is 3.45 times larger than total membrane area of the 3-tube PMR (0.0247 m{sup 2}). The following objectives were identified for the 7-tube PMR tests: (1) Refine test measurements, especially humidity and flow; (2) Refine maintenance procedures for Pd/Ag tube conditioning; (3) Evaluate baseline PMR operating conditions; (4) Determine PMR

  8. Fabrication of U-10 wt.%Zr Metallic Fuel Rodlets for Irradiation Test in BOR-60 Fast Reactor

    OpenAIRE

    Ki-Hwan Kim; Jong-Hwan Kim; Seok-Jin Oh; Jung-Won Lee; Ho-Jin Lee; Chan-Bock Lee

    2016-01-01

    The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr...

  9. A test study on treatment of high-strength polyester wastewater with anaerobic reactor

    Institute of Scientific and Technical Information of China (English)

    韩洪军; 陈秀荣; 徐春艳

    2002-01-01

    The treatment of polyester wastewater using Up-flow activated sludge bed anaerobic filer ( UASB-AF), demonstrated that UASB-AF reactors has a high efficiency, its volume loading is 10 ~ 12 kgCOD/( m3 @d) ,HRT is 22 ~24 h, and the removal of COD is about 80%. The reactor has advantage of fast starting andenduring pulse loading.

  10. Thermal hydraulics in the hot pool of Fast Breeder Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Padmakumar, G. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India)], E-mail: gpk@igcar.gov.in; Pandey, G.K.; Vaidyanathan, G. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India)

    2009-06-15

    Sodium cooled Fast Breeder Test Reactor (FBTR) of 40 MWt/13 MWe capacity is in operation at Kalpakkam, near Chennai. Presently it is operating with a core of 10.5 MWt. Knowledge of temperatures and flow pattern in the hot pool of FBTR is essential to assess the thermal stresses in the hot pool. While theoretical analysis of the hot pool has been conducted by a three-dimensional code to access the temperature profile, it involves tuning due to complex geometry, thermal stresses and vibration. With this in view, an experimental model was fabricated in 1/4 scale using acrylic material and tests were conducted in water. Initially hydraulic studies were conducted with ambient water maintaining Froude number similarity. After that thermal studies were conducted using hot and cold water maintaining Richardson similitude. In both cases Euler similarity was also maintained. Studies were conducted simulating both low and full power operating conditions. This paper discusses the model simulation, similarity criteria, the various thermal hydraulic studies that were carried out, the results obtained and the comparison with the prototype measurements.

  11. Laboratory test reactor for the investigation of liquid reducing agents in the selective catalytic reduction of NOx

    Science.gov (United States)

    Peitz, D.; Bernhard, A.; Elsener, M.; Kröcher, O.

    2011-08-01

    A test reactor was designed and built for investigating liquid reducing agents in the selective catalytic reduction (SCR) process in the laboratory. The design of the experimental setup is described in detail and its performance was evaluated. Using a glass nebulizer, liquid reducing agents were sprayed directly onto a catalyst positioned in a heated glass reactor with a length of 250 mm and an internal diameter of 20.4 mm or 40 mm. Model exhaust gases were mixed from individual gas components and were heated up to 450 °C in a heat exchanger before entering the reactor. The off-gas was analyzed using two complimentary techniques, a multi-component online FTIR gas analysis and a liquid quench gas absorption setup, to detect higher molecular compounds and aerosols. Due to the versatility of construction, processes not related to SCR, but involving three-phase reactions with gases, liquids and a catalyst, can also be investigated.

  12. Application of automatic inspection system to nondestructive test of heat transfer tubes of primary pressurized water cooler in the high temperature engineering test reactor. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Furusawa, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Miyamoto, Satoshi [Japan Atomic Power Company, Tokyo (Japan)

    2001-07-01

    Heat transfer tubes of a primary pressurized water cooled (PPWC) in the high temperature engineering test reactor (HTTR) form the reactor pressure boundary of the primary coolant, therefore are important from the viewpoint of safety. To establish inspection techniques for the heat transfer tubes of the PPWC, an automatic inspection system was developed. The system employs a bobbin coil probe, a rotating probe for eddy current testing (ECT) and a rotating probe for ultrasonic testing (UT). Nondestructive test of a half of the heat transfer tubes of the PPWC was carried out by the automatic inspection system during reactor shutdown period of the HTTR (about 55% in the maximum reactor power in this paper). The nondestructive test results showed that the maximum signal-to-noise ratio was 1.8 in ECT. Pattern and phase of Lissajous wave, which were obtained for the heat transfer tube of the PPWC, were different from those obtained for the artificially defected tube. In UT echo amplitude of the PPWC tubes inspected was lower than 20% of distance-amplitude calibration curve. Thus, it was confirmed that there was no defect in depth, which was more than the detecting standard of the probes, on the outer surface of the heat transfer tubes of the PPWC inspected. (author)

  13. Parametric experimental tests of steam gasification of pine wood in a fluidized bed reactor

    Directory of Open Access Journals (Sweden)

    L. Vecchione

    2013-09-01

    Full Text Available Among Renewable Energy Sources (RES, biomass represent one of the most common and suitable solution in order to contribute to the global energy supply and to reduce greenhouse gases (GHG emissions. The disposal of some residual biomass, as pruning from pine trees, represent a problem for agricultural and agro-industrial sectors. But if the residual biomass are used for energy production can become a resource. The most suitable energy conversion technology for the above-mentioned biomass is gasification process because the high C/N ratio and the low moisture content, obtained from the analysis. In this work a small-pilot bubbling-bed gasification plant has been designed, constructed and used in order to obtain, from the pine trees pruning, a syngas with low tar and char contents and high hydrogen content. The activities showed here are part of the activities carried out in the European 7FP UNIfHY project. In particular the aim of this work is to develop experimental test on a bench scale steam blown fluidized bed biomass gasifier. These tests will be utilized in future works for the simulations of a pilot scale steam fluidized bed gasifier (100 kWth fed with different biomass feedstock. The results of the tests include produced gas and tar composition as well gas, tar and char yield. Tests on a bench scale reactor (8 cm I.D. were carried out varying steam to biomass ratio from 0.5, 0.7 and 1 to 830°C.

  14. Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

    Energy Technology Data Exchange (ETDEWEB)

    Morris, Robert N., E-mail: morrisrn@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Baldwin, Charles A. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Demkowicz, Paul A. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Hunn, John D. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Reber, Edward L. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2016-09-15

    Highlights: • High-temperature safety tests were performed on 14 irradiated HTGR fuel compacts. • Significant krypton release was detected in only one of the safety tests. • Cesium retention by intact SiC was excellent, even up to 1800 °C. • Release of Ag, Eu, and Sr was dominated by previous release during irradiation. • Silver exhibited the highest fractional release. - Abstract: The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ∼58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products {sup 110m}Ag, {sup 134}Cs, {sup 137}Cs, {sup 154}Eu, {sup 155}Eu, {sup 90}Sr, and {sup 85}Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ∼100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10{sup −6} after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and {sup

  15. TEMPERATURE MONITORING OPTIONS AVAILABLE AT THE IDAHO NATIONAL LABORATORY ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    J.E. Daw; J.L. Rempe; D.L. Knudson; T. Unruh; B.M. Chase; K.L Davis

    2012-03-01

    As part of the Advanced Test Reactor National Scientific User Facility (ATR NSUF) program, the Idaho National Laboratory (INL) has developed in-house capabilities to fabricate, test, and qualify new and enhanced sensors for irradiation testing. To meet recent customer requests, an array of temperature monitoring options is now available to ATR users. The method selected is determined by test requirements and budget. Melt wires are the simplest and least expensive option for monitoring temperature. INL has recently verified the melting temperature of a collection of materials with melt temperatures ranging from 100 to 1000 C with a differential scanning calorimeter installed at INL’s High Temperature Test Laboratory (HTTL). INL encapsulates these melt wires in quartz or metal tubes. In the case of quartz tubes, multiple wires can be encapsulated in a single 1.6 mm diameter tube. The second option available to ATR users is a silicon carbide temperature monitor. The benefit of this option is that a single small monitor (typically 1 mm x 1 mm x 10 mm or 1 mm diameter x 10 mm length) can be used to detect peak irradiation temperatures ranging from 200 to 800 C. Equipment has been installed at INL’s HTTL to complete post-irradiation resistivity measurements on SiC monitors, a technique that has been found to yield the most accurate temperatures from these monitors. For instrumented tests, thermocouples may be used. In addition to Type-K and Type-N thermocouples, a High Temperature Irradiation Resistant ThermoCouple (HTIR-TC) was developed at the HTTL that contains commercially-available doped molybdenum paired with a niobium alloy thermoelements. Long duration high temperature tests, in furnaces and in the ATR and other MTRs, demonstrate that the HTIR-TC is accurate up to 1800 C and insensitive to thermal neutron interactions. Thus, degradation observed at temperatures above 1100 C with Type K and N thermocouples and decalibration due to transmutation with tungsten

  16. Manufacturing and testing in reactor relevant conditions of brazed plasma facing components of the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Bisio, M. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Branca, V. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Marco, M. Di [FN s.p.a., ss 35 bis dei Giovi km 15, I-15062 Bosco Marengo (Albania) (Italy); Federici, A. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Grattarola, M. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy)]. E-mail: grattarola@ansaldo.it; Gualco, G. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Guarnone, P. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Luconi, U. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Merola, M. [EFDA, Boltzmanstr. 2, D-85748 Garching (Germany); Ozzano, C. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Pasquale, G. [FN s.p.a., ss 35 bis dei Giovi km 15, I-15062 Bosco Marengo (AL) (Italy); Poggi, P. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Rizzo, S. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy); Varone, F. [Ansaldo Ricerche s.p.a., C.so Perrone 25, I-16152 Genova (Italy)

    2005-11-15

    A fabrication route based on brazing technology has been developed for the realization of the high heat flux components for the ITER vertical target and Dome-Liner. The divertor vertical target is armoured with carbon fiber reinforced carbon and tungsten in the lower straight part and in the upper curved part, respectively. The armour material is joined to heat sinks made of precipitation hardened copper-chromium-zirconium alloy. The plasma facing units of the dome component are based on a tungsten flat tile design with hypervapotron cooling. An innovative brazing technique based on the addition of carbon fibers to the active brazing alloy, developed by Ansaldo Ricerche for applications in the field of the energy production, has been used for the carbon fiber composite to copper joint to reduce residual stresses. The tungsten-copper joint has been realized by direct casting. A proper brazing thermal cycle has been studied to guarantee the required mechanical properties of the precipitation hardened alloy after brazing. The fabrication route of plasma facing components for the ITER vertical target and dome based on the brazing technology has been proved by means of thermal fatigue tests performed on mock-ups in reactor relevant conditions.

  17. Activation calculation and radiation analysis for China Fusion Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Zhi, E-mail: zchen@ustc.edu.cn; Qiao, Shiji; Jiang, Shuai; Xu, X. George

    2016-11-01

    Highlights: • Activation calculation was performed using FLUKA for the main components of CFETR. • Radionuclides and radioactive wastes were assessed for CFETR. • The Waste Disposal Ratings (WDR) were assessed for CFETR. - Abstract: The activation calculation and analysis for the China Fusion Engineering Test Reactor (CFETR) will play an important role in its system design, maintenance, inspection and assessment of nuclear waste. Using the multi-particle transport code FLUKA and its associated data library, we calculated the radioactivity, specific activity, waste disposal rating from activation products, nuclides in the tritium breeding blanket, shielding layer, vacuum vessel and toroidal field coil (TFC) of CFETR. This paper presents the calculation results including neutron flux, activation products and waste disposal rating after one-year full operation of the CFETR. The findings show that, under the assumption of one-year operation at the 200 MW fusion power, the total radioactivity inventory will be 1.05 × 10{sup 19} Bq at shutdown and 1.03 × 10{sup 17} Bq after ten years. The primary residual nuclide is found to be {sup 55}Fe in ten years after the shutdown. The waste disposal rating (WDR) values are very low (<<1), according to Class C limits, CFETR materials are qualified for shallow land burial. It is shown that CFETR has no serious activation safety issue.

  18. Anomalous fast ion losses at high β on the tokamak fusion test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fredrickson, E. D.; Bell, M. G.; Budny, R. V.; Darrow, D. S.; White, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

    2015-03-15

    This paper describes experiments carried out on the Tokamak Fusion Test Reactor (TFTR) [R. J. Hawryluk et al., Plasma Phys. Controlled Fusion 33, 1509 (1991)] to investigate the dependence of β-limiting disruption characteristics on toroidal field strength. The hard disruptions found at the β-limit in high field plasmas were not found at low field, even for β's 50% higher than the empirical β-limit of β{sub n} ≈ 2 at high field. Comparisons of experimentally measured β's to TRANSP simulations suggest anomalous loss of up to half of the beam fast ions in the highest β, low field shots. The anomalous transport responsible for the fast ion losses may at the same time broaden the pressure profile. Toroidal Alfvén eigenmodes, fishbone instabilities, and Geodesic Acoustic Modes are investigated as possible causes of the enhanced losses. Here, we present the first observations of high frequency fishbones [F. Zonca et al., Nucl. Fusion 49, 085009 (2009)] on TFTR. The interpretation of Axi-symmetric Beam-driven Modes as Geodesic Acoustic Modes and their possible correlation with transport barrier formation are also presented.

  19. Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

    2012-02-01

    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 (NGNP) project. Because the NRC's 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).

  20. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014

    Energy Technology Data Exchange (ETDEWEB)

    Soelberg, Renae [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-11-01

    Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014 Highlights Rory Kennedy and Sarah Robertson attended the American Nuclear Society Winter Meeting and Nuclear Technology Expo in Anaheim, California, Nov. 10-13. ATR NSUF exhibited at the technology expo where hundreds of meeting participants had an opportunity to learn more about ATR NSUF. Dr. Kennedy briefed the Nuclear Engineering Department Heads Organization (NEDHO) on the workings of the ATR NSUF. • Rory Kennedy, James Cole and Dan Ogden participated in a reactor instrumentation discussion with Jean-Francois Villard and Christopher Destouches of CEA and several members of the INL staff. • ATR NSUF received approval from the NE-20 office to start planning the annual Users Meeting. The meeting will be held at INL, June 22-25. • Mike Worley, director of the Office of Innovative Nuclear Research (NE-42), visited INL Nov. 4-5. Milestones Completed • Recommendations for the Summer Rapid Turnaround Experiment awards were submitted to DOE-HQ Nov. 12 (Level 2 milestone due Nov. 30). Major Accomplishments/Activities • The University of California, Santa Barbara 2 experiment was unloaded from the GE-2000 at HFEF. The experiment specimen packs will be removed and shipped to ORNL for PIE. • The Terrani experiment, one of three FY 2014 new awards, was completed utilizing the Advanced Photon Source MRCAT beamline. The experiment investigated the chemical state of Ag and Pd in SiC shell of irradiated TRISO particles via X-ray Absorption Fine Structure (XAFS) spectroscopy. Upcoming Meetings/Events • The ATR NSUF program review meeting will be held Dec. 9-10 at L’Enfant Plaza. In addition to NSUF staff and users, NE-4, NE-5 and NE-7 representatives will attend the meeting. Awarded Research Projects Boise State University Rapid Turnaround Experiments (14-485 and 14-486) Nanoindentation and TEM work on the T91, HT9, HCM12A and 9Cr ODS specimens has been completed at

  1. Thermal analysis for a spent reactor fuel storage test in granite

    Energy Technology Data Exchange (ETDEWEB)

    Montan, D.N.

    1980-09-01

    A test is conducted in which spent fuel assemblies from an operating commercial nuclear power reactor are emplaced in the Climax granite at the US Department of Energy`s Nevada Test Site. In this generic test, 11 canisters of spent PWR fuel are emplaced vertically along with 6 electrical simulator canisters on 3 m centers, 4 m below the floor of a storage drift which is 420 m below the surface. Two adjacent parallel drifts contain electrical heaters, operated to simulate (in the vicinity of the storage drift) the temperature fields of a large repository. This test, planned for up to five years duration, uses fairly young fuel (2.5 years out of core) so that the thermal peak will occur during the time frame of the test and will not exceed the peak that would not occur until about 40 years of storage had older fuel (5 to 15 years out of core) been used. This paper describes the calculational techniques and summarizes the results of a large number of thermal calculations used in the concept, basic design and final design of the spent fuel test. The results of the preliminary calculations show the effects of spacing and spent fuel age. Either radiation or convection is sufficient to make the drifts much better thermal conductors than the rock that was removed to create them. The combination of radiation and convection causes the drift surfaces to be nearly isothermal even though the heat source is below the floor. With a nominal ventilation rate of 2 m{sup 3}/s and an ambient rock temperature of 23{sup 0}C, the maximum calculated rock temperature (near the center of the heat source) is about 100{sup 0}C while the maximum air temperature in the drift is around 40{sup 0}C. This ventilation (1 m{sup 3}/s through the main drift and 1/2 m{sup 3}/s through each of the side drifts) will remove about 1/3 of the heat generated during the first five years of storage.

  2. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Devin A. Steuhm

    2011-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose. Furthermore

  3. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg, Principal Investigator; Kevin A. Steuhm, Project Manager

    2012-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core

  4. Long term out-of-pile thermocouple tests in conditions representative for nuclear gas-cooled high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Laurie, M., E-mail: mathias.laurie@ec.europa.eu [European Commission, Joint Research Centre, Institut für Transurane, Hermann-von-Helmholtz-Platz 1, 76344, Eggenstein-Leopoldshafen (Germany); Fourrez, S. [THERMOCOAX SAS, Rue du Pré Neuf, 61100 Saint Georges des Groseillers (France); Fütterer, M.A.; Lapetite, J.M. [European Commission, Joint Research Centre, Institut für Transurane, Hermann-von-Helmholtz-Platz 1, 76344, Eggenstein-Leopoldshafen (Germany); Sadli, M.; Morice, R.; Failleau, G. [Laboratoire commun de métrologie LNE-Cnam, 61 rue du Landy, F-92310 Saint-Denis (France)

    2014-05-01

    During irradiation tests at high temperature failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. As instrumentation, in particular thermocouples are considered safety-relevant both for irradiation tests and for commercial reactors, JRC and THERMOCOAX joined forces to solve this issue by performing out-of-pile tests with thermocouples mimicking the environment encountered by high temperature reactor (HTR) in-core instrumentation. The objective was to screen innovative sheathed thermocouples which would consecutively be tested under irradiation. Two such screening tests have been performed in high temperature environment (i.e. temperature in the range 1100–1150 °C) with purposely contaminated helium atmosphere (mainly CH{sub 4}, CO, CO{sub 2}, O{sub 2} impurities) representative for high temperature reactor carburizing atmospheres. The first set of thermocouples embedded in graphite (mainly conventional N type thermocouples and thermocouples with innovative sheaths) was tested in a dedicated furnace at THERMOCOAX lab with helium flushing. The second out-of-pile test at JRC with a partly different set of thermocouples replicated the original test for comparison. Performance indicators such as thermal drift, insulation resistance and loop resistance were monitored. Through these long-term screening tests the effect of several parameters were investigated: niobium sleeves, bending, diameter, sheath composition as well as the chemical environment. SEM examinations were performed to analyze local damage (bending zone, sheath). The present paper describes the two tests, sums up data collected during these tests in terms of thermocouple behavior and describes further instrumentation testing work with fixed point mini cells for qualification under irradiation.

  5. Small Punch Test on Before and Post Irradiated Domestic Reactor Pressure Steel

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Problems may be caused when applying the standard specimen to study the properties of irradiated reactor materials, because of its big dimension, e.g.: The inner temperature gradient of the specimen is high when irradiated, the radiation

  6. CO2 Absorption in a Lab-Scale Fixed Solid Bed Reactor: Modelling and Experimental Tests

    Directory of Open Access Journals (Sweden)

    Roberto Gabbrielli

    2004-09-01

    Full Text Available The CO2 absorption in a lab-scale fixed solid bed reactor filled with different solid sorbents has been studied under different operative conditions regarding temperature (20-200°C and input gas composition (N2, O2, CO2, H2O at 1bar pressure. The gas leaving the reactor has been analysed to measure the CO2 and O2 concentrations and, consequently, to evaluate the overall CO2 removal efficiency. In order to study the influence of solid sorbent type (i.e. CaO, coal bottom ash, limestone and blast furnace slag and of mass and heat transfer processes on CO2 removal efficiency, a one-dimensional time dependent mathematical model of the reactor, which may be considered a Plug Flow Reactor, has been developed. The quality of the model has been confirmed using the experimental results.

  7. ENVIRONMENTAL TECHNOLOGY VERIFICATION REPORT, REMOVAL OF ARSENIC IN DRINKING WATER: WATTS PREMIER M-SERIES M-15,000 REVERSE OSMOSIS TREATMENT SYSTEM

    Science.gov (United States)

    Verification testing of the Watts Premier M-Series M-15,000 RO Treatment System was conducted over a 31-day period from April 26, 2004, through May 26, 2004. This test was conducted at the Coachella Valley Water District (CVWD) Well 7802 in Thermal, California. The source water...

  8. ENVIRONMENTAL TECHNOLOGY VERIFICATION REPORT, REMOVAL OF ARSENIC IN DRINKING WATER: WATTS PREMIER M-SERIES M-15,000 REVERSE OSMOSIS TREATMENT SYSTEM

    Science.gov (United States)

    Verification testing of the Watts Premier M-Series M-15,000 RO Treatment System was conducted over a 31-day period from April 26, 2004, through May 26, 2004. This test was conducted at the Coachella Valley Water District (CVWD) Well 7802 in Thermal, California. The source water...

  9. EVALUATION OF ZERO-POWER, ELEVATED-TEMPERATURE MEASUREMENTS AT JAPAN’S HIGH TEMPERATURE ENGINEERING TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

    2011-03-01

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The experimental benchmark performed and currently evaluated in this report pertains to the data available for two zero-power, warm-critical measurements with the fully-loaded HTTR core. Six isothermal temperature coefficients for the fully-loaded core from approximately 340 to 740 K have also been evaluated. These experiments were performed as part of the power-up tests (References 1 and 2). Evaluation of the start-up core physics tests specific to the fully-loaded core (HTTR-GCR-RESR-001) and annular start-up core loadings (HTTR-GCR-RESR-002) have been previously evaluated.

  10. Fabrication of U-10 wt.%Zr Metallic Fuel Rodlets for Irradiation Test in BOR-60 Fast Reactor

    Directory of Open Access Journals (Sweden)

    Ki-Hwan Kim

    2016-01-01

    Full Text Available The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr fuel slugs with a diameter of 5.5 mm. Consequently, fuel slugs per melting batch without casting defects were fabricated through the development of advanced casting technology and evaluation tests. The optimal GTAW welding conditions were also established through a number of experiments. In addition, a qualification test was carried out to prove the weld quality of the end plug welding of the metallic fuel rodlets. The wire wrapping of metallic fuel rodlets was successfully accomplished for the irradiation test. Thus, PGSFR fuel rodlets have been soundly fabricated for the irradiation test in a BOR-60 fast reactor.

  11. Vice Premier Li Keqiang Meets Chinese and Russian Friendly Personages

    Institute of Scientific and Technical Information of China (English)

    2012-01-01

    <正>The scene: 5 pm on April 26 at the President Hotel in Moscow, and Vice Premier Li Keqiang, having just arrived in Russia for an official visit, cordially meets Chinese and Russian friendly personages from various sectors of society including people-to-people diplomatic, academic, cultural, scientific and educational, and business circles. Among them are experts and scholars in their seventies and eighties, important professionals of their respective workplaces in their prime, as well as students in the bloom of their youth.

  12. Lessons premier hospitals learned about implementing electronic health records.

    Science.gov (United States)

    DeVore, Susan D; Figlioli, Keith

    2010-04-01

    Implementing health information technology (IT) is a major strategic objective for providers. To pinpoint considerations that tie to success, the Premier health care alliance surveyed hospitals to develop an electronic health record best-practices library. Compiled from diverse health care organizations, the library outlines considerations to support "meaningful use" in the areas of computerized physician order entry, medication management, clinical documentation, reporting of measures, privacy, information exchange, management of populations' health, and personal health records. Best practices also uncovered strategies for securing executive leadership, culture change, communication, and support for clinicians. This paper summarizes lessons from the library, providing recommendations to speed up health IT implementation.

  13. Safety of power transformers, power supplies, reactors and similar products - Part 1: General requirements and tests

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1998-01-01

    This International Standard deals with safety aspects of power transformers, power supplies, reactors and similar products such as electrical, thermal and mechanical safety. This standard covers the following types of dry-type transformers, power supplies, including switch mode power supplies, and reactors, the windings of which may be encapsulated or non-encapsulated. It has the status of a group safety publication in accordance with IEC Guide 104.

  14. External Attachment of Titanium Sheathed Thermocouples to Zirconium Nuclear Fuel Rods For The Loss-Of-Fluid-Test (LOFT) Reactor

    Science.gov (United States)

    Welty, Richard K.

    1980-10-01

    The Exxon Nuclear Company, Inc. acting as a Subcontractor to EG&G Idaho Inc.3 Idaho National Engineering Laboratory, Idaho Falls, Idaho, has developed a welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods. The fuel rods and thermocouples are used to test simulated loss-of-coolant-accident (LOCA) conditions in a pressurized water reactor (LOFT Reactor, Idaho National Laboratory). The design goals were to (1) reliably attach thermocouples to the zircaloy fuel rods, (2) achieve or exceed a life expectancy of 6,000 hours of reactor operation in a borated water environment of 316°C at 2260 psi, (3) provide and sustain repeatable physical and metallurgical properties in the instrumented rods subjected to transient temperatures up to 1538°C with blowdown, shock, loading, and fast quench. A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A commercial pulsed laser and energy control system was installed along with specialized welding fixtures. Laser room facility requirements and tolerances were established. Performance qualifications and detailed welding procedures were also developed. Product performance tests were conducted to assure that engineering design requirements could be met on a production basis. Irradiation tests showed no degradation of thermocouples or weld structure. Fast thermal cycle and heater rod blowdown reflood tests were made to subject the weldments to high temperatures, high pressure steam, and fast water quench cycles. From the behavior of these tests, it was concluded that the attachment welds would survive a series of reactor safety tests.

  15. Post test calculations of a severe accident experiment for VVER-440 reactors by the ATHLET code

    Energy Technology Data Exchange (ETDEWEB)

    Gyoergy, Hunor [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques (BME NTI); Trosztel, Istvan [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research (MTA EK)

    2013-09-15

    Severe accident - if no mitigation action is taken - leads to core melt. An effective severe accident management strategy can be the external reactor pressure vessel cooling for corium localization and stabilization. For some time discussion was going on, whether the in-vessel retention can be applied for the VVER-440 type reactors. It had to be demonstrated that the available space between the reactor vessel and biological protection allows sufficient cooling to keep the melted core in the vessel, without the reactor pressure vessel losing its integrity. In order to demonstrate the feasibility of the concept an experimental facility was realized in Hungary. The facility called Cooling Effectiveness on the Reactor External Surface (CERES) is modeling the vessel external surface and the biological protection of Paks NPP. A model of the CERES facility for the ATHLET TH system code was developed. The results of the ATHLET calculation agree well with the measurements showing that the vessel cooling can be insured for a long time in a VVER-440 reactor. (orig.)

  16. Turbulent jet erosion of a stably stratified gas layer in a nuclear reactor test containment

    Energy Technology Data Exchange (ETDEWEB)

    Ishay, Liel [Department of Mechanical Engineering, Ben-Gurion University of the Negev, Beer-Sheva 84105 (Israel); Bieder, Ulrich [Commissariat à l’énergie atomique et aux énergies alternatives, Centre de SACLAY DEN/SAC/DANS/DM2S/STMF/LMSF, F-91191 Gif-sur-Yvette (France); Ziskind, Gennady [Department of Mechanical Engineering, Ben-Gurion University of the Negev, Beer-Sheva 84105 (Israel); Rashkovan, Alex, E-mail: rashbgu@gmail.com [Physics Department, Nuclear Research Center Negev (NRCN), PO Box 9001, Beer-Sheva 84190 (Israel)

    2015-10-15

    Highlights: • We model stably stratified layer erosion by vertical turbulent round jet. • Separate effect studies are performed as a platform for choosing modeling approach. • A test performed in MISTRA facility, CEA, Saclay is modeled using Fluent and Trio-U codes. • The proposed modeling approach showed good agreement with the MISTRA facility LOWMA-3 test. - Abstract: A number of integral and separate effect experiments were performed in the last two decades for validation of containment computational tools. The main goal of these benchmark experiments was to assess the ability of turbulence models and computational fluid dynamics codes to predict hydrogen concentration distribution and steam condensation rate in a nuclear reactor containment in the course of severe accidents. It appears from the published literature that the predictive capability of the existing computational tools still needs to be improved. This work examines numerically the temporal evolution of helium concentration in the experiment called LOWMA-3, performed in the MISTRA facility of CEA-Saclay, France. In the experiment, helium is used to mimic hydrogen of a real-case accident. The aim of this separate effect experiment, where steam condensation was not involved, is to predict helium concentration field. The conditions of the experiment are such that both the momentum transport and molecular diffusion contributions to the mixing process are of the same order of magnitude (Fr ∼ 1). A commercial CFD code, Fluent, and a CEA in-house code, Trio-U, are used for flow and helium concentration fields temporal evolution prediction in the present study. The preliminary separate effect studies provide guidance to an optimal modeling approach for the LOWMA-3 experiment. Temporal evolution of helium concentration in the stratification layer is shown, and a comparison to the experiment is discussed. It is shown that correct modeling of the round jet flowfield is essential for a reliable

  17. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  18. Modelling of turbulent hydrocarbon combustion. Test of different reactor concepts for describing the interactions between turbulence and chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.; Kremer, H. [Ruhr-Universitaet Bochum, Lehrstuhl fuer Energieanlagentechnik, Bochum (Germany); Kilpinen, P.; Hupa, M. [Aabo Akademi, Turku (Finland). Combustion Chemistry Research Group

    1997-12-31

    The detailed modelling of turbulent reactive flows with CFD-codes is a major challenge in combustion science. One method of combining highly developed turbulence models and detailed chemistry in CFD-codes is the application of reactor based turbulence chemistry interaction models. In this work the influence of different reactor concepts on methane and NO{sub x} chemistry in turbulent reactive flows was investigated. Besides the classical reactor approaches, a plug flow reactor (PFR) and a perfectly stirred reactor (PSR), the Eddy-Dissipation Combustion Model (EDX) and the Eddy Dissipation Concept (EDC) were included. Based on a detailed reaction scheme and a simplified 2-step mechanism studies were performed in a simplified computational grid consisting of 5 cells. The investigations cover a temperature range from 1273 K to 1673 K and consider fuel-rich and fuel-lean gas mixtures as well as turbulent and highly turbulent flow conditions. All test cases investigated in this study showed a strong influence of the reactor residence time on the species conversion processes. Due to this characteristic strong deviations were found for the species trends resulting from the different reactor approaches. However, this influence was only concentrated on the `near burner region` and after 4-5 cells hardly any deviation and residence time dependence could be found. The importance of the residence time dependence increased when the species conversion was accelerated as it is the case for overstoichiometric combustion conditions and increased temperatures. The study focused furthermore on the fine structure in the EDC. Unlike the classical approach this part of the cell was modelled as a PFR instead of a PSR. For high temperature conditions there was hardly any difference between both reactor types. However, decreasing the temperature led to obvious deviations. Finally, the effect of the selective species transport between the cells on the conversion process was investigated

  19. RANKING THE SPECTATORS’ DIFFICULTIES IN PURCHASING ELECTRONIC TICKETS OF FOOTBALL PREMIER LEAGUE

    Directory of Open Access Journals (Sweden)

    Ahmad Narimani

    2017-04-01

    Full Text Available This study aimed to rank the spectators’ difficulties in buying electronic tickets of football premier league matches at Azadi stadium. The population consisted of all spectators of Esteghlal-Persepolis match in the fifteenth league at Azadi stadium (N= 100000. According to Morgan table and using simple random sampling method, 500 participants were selected as sample. A researcher-made questionnaire was used for collecting the data; its face validity was confirmed by 15 experts and performing a pilot study on 30 subjects, its Cronbach’s alpha was calculated to be 0.86. Using SPSS 22, the descriptive and inferential (including Friedman test statistics was applied for analyzing the data. The findings showed that there was a significant difference between rankings of difficulties in buying electronic tickets of Football premier league matches at Azadi Stadium. The difficulties were ranked as: problem in ticket systems, early selling out of electronic tickets, lack of confidence to electronic ticket sale, lack of skill to work with the internet, low speed of internet, and lack of access to the internet

  20. The Relationship between Emotional Intelligence and Job Satisfaction among Coaches in Premier Under-20 Football League

    Directory of Open Access Journals (Sweden)

    Mehdi Moradi

    2012-06-01

    Full Text Available The purpose of present study was to examine the relationship between emotional intelligence and job satisfaction among coaches in premier Under-20 football league. The research method was descriptive-correlative, the performance method was survey, and data collection was done through field study. Research population consisted of 56 coaching staff in 14 teams participating in premier Under-20 football league. Finally, there were 48 questionnaires useable in data analysis. Emotional Intelligence Questionnaire (Syber Yashring and JDI (Wysocki & Kromm were used to collect the data. Descriptive statistics was used to describe data, Kolmogorov-Smirnov test was used to know whether the distribution of data was normal, and Pearson correlation and stepwise regression were applied to investigate the significance of hypotheses. Results showed that there was significant association between emotional intelligence, subscale self-awareness, subscale empathy, and subscale social skills with job satisfaction (p≤0.05. However, there was not significant association between subscale self-motivation and subscale self-control with job satisfaction. Self-awareness, empathy, and social skills (predictors predicted job satisfaction (criterion significantly. Predicted value of self-awareness, empathy, and social skills was 0.4, 0.29, and 0.26 respectively. Training and aging increase emotional intelligence so it is predicted more job satisfaction over the time. From other side, clubs and football federation as the head can create scientific atmosphere and instruct psychological coaching principles. It will lead to enjoy creative, willing players as output.

  1. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B.

    1997-04-01

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

  2. Analytical evaluation on loss of off-side electric power simulation of the High Temperature Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Nakagawa, Shigeaki; Tachibana, Yukio; Takada, Eiji; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2000-03-01

    A rise-to-power test of the high temperature engineering test reactor (HTTR) started on September 28 in 1999 for establishing and upgrading the technological basis for the high temperature gas-cooled reactor (HTGR). A loss of off-site electric power test of the HTTR from the normal operation under 15 and 30 MW thermal power will be carried out in the rise-to-power test. Analytical evaluations on transient behaviors of the reactor and plant during the loss of off-site electric power were conducted. These estimations are proposed as benchmark problems for the IAEA coordinated research program on 'Evaluation of HTGR Performance'. This report describes an event scenario of transient during the loss of off-site electric power, the outline of major components and system, detailed thermal and nuclear data set for these problems and pre-estimation results of the benchmark problems by an analytical code 'ACCORD' for incore and plant dynamics of the HTGR. (author)

  3. Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Forsbacka, Matthew

    2004-01-01

    For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the NASA Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. "virtual" reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of van'ous core deformations. The power delivered to the SAFE-100 prototype was then dusted accordingly via kinetics calculations, The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kilowatt(sub t), to 10 kilowatts(sub t), held approximately constant at 10 kilowatts (sub t), and then allowed to decrease based on the negative thermal reactivity coefficient.

  4. Characterization of spent fuel elements stored at IEA-R1 research reactor based on visual inspections and sipping tests

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Jose Eduardo Rosa da; Terremoto, Luis Antonio Albiac; Teodoro, Celso Antonio; Castanheira, Myrthes; Lucki, Georgi; Damy, Margaret de Almeida; Silva, Antonio Teixeira e [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: jersilva@ipen.br

    2005-07-01

    Aluminum spent nuclear fuels are susceptible to corrosion attack, or mechanical damage from improper handling, while in pool reactor storage. Storage practices have been modified to reduce the potential for damage, based on recommendations presented at second WS on Spent Fuel Characterization, promoted by IAEA. In this work, we present the inspection program proposed to the IEA-R1 stored spent fuel elements, in order to provide information on the physical condition during the interim storage time under wet condition at the reactor pool. The inspection program is based on non-destructive tests results (visual inspection and sipping tests) already periodically performed to exam the IEA-R1 stored spent fuel and fuel elements from the core reactor. To record the available information and examination results it was elaborated a document in the format of a catalogue containing the proposed inspection program for the IEA-R1 stored spent fuel, the description of the visual inspection and sipping tests systems, a compilation of information and images result from the tests performed for all stored standard spent fuel element and, in annexes, copies of the reference documents. That document constitutes an important step of the effective implementation of the referred IEA-R1 spent fuel inspection program and can be used to address regulatory and operational needs for the demonstration, for example, of safe storage throughout the pool storage period. (author)

  5. Development and testing of a pilot reactor for enzymatic hydrolysis of cellulose-containing waste. Entwicklung und Erprobung eines Pilotreaktors fuer die enzymatische Hydrolyse cellulosehaltiger Abfallstoffe

    Energy Technology Data Exchange (ETDEWEB)

    Borchert, A.

    1985-08-01

    Enzymatic hydrolysis of cellulose- and pectin-containing waste materials requires simple reactors and a simple, adaptable technology. The conversion process was studied with a view to reactor design, using a discontinuous trickle-bed reactor and a continuous spiral-shaft reactor, each with a volume of 50 l. Both reactors are based on the tube reactor principle, which had already been tested on a small scale in an earlier project. The substrates were potato pulp, shredded beetroot, and pretreated wheat straw. Enzyme mixtures for these substrates were chosen with a view to achieving good yields within 48 h at the most. The main parameters were the time of residue, temperature, enzyme adsorption and stability, and an experimental procedure with or without partial recirculation of the reaction products.

  6. Nuclear reactor (1960); Reacteurs nucleaires (1960)

    Energy Technology Data Exchange (ETDEWEB)

    Maillard, M.L. [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires; Leo, M.B. [Electricite de France (EDF), 75 - Paris (France)

    1960-07-01

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [French] Les premiers reacteurs industriels plutonigenes francais G1 - G2 - G3 du Centre de Marcoule comportent une installation de recuperation d'energie. La production d'electricite de G1 ne compense pas l'energie depensee par ailleurs pour le fonctionnement de l'ensemble, par contre, G2 et G3 doivent fournir chacun une puissance de 25 a 30 MW au reseau national d'Electricite de France. Cette puissance est modeste, mais l'experience acquise grace a ces reacteurs est tres grande et c'est grace a elle qu'il nous sera possible de mettre en exploitation les reacteurs energetiques EDF1 - EDF2 - EDF3. Le memoire decrit comment, avant tout demarrage du reacteur, les essais effectues, en particulier ceux concernant l'installation de recuperation d'energie et le caisson, ont permis d'abreger la phase de montee en puissance. (auteur)

  7. Investigation of TASS/SMR Capability to Predict a Natural Circulation in the Test Facility for an Integral Reactor

    Directory of Open Access Journals (Sweden)

    Young-Jong Chung

    2014-01-01

    Full Text Available System-integrated modular advanced reactor (SMART is a small-sized advanced integral type pressurized water reactor (PWR with a rated thermal power of 330 MW. It can produce 100 MW of electricity or 90 MW of electricity and 40,000 ton of desalinated water concurrently, which is sufficient for 100,000 residents. The design features contributing to safety enhancement are basically inherent safety improvement and passive safety features. TASS/SMR code was developed for an analysis of design based events and accidents in an integral type reactor reflecting the characteristics of the SMART design. The main purpose of the code is to analyze all relevant phenomena and processes. The code should be validated using experimental data in order to confirm prediction capability. TASS/SMR predicts well the overall thermal-hydraulic behavior under various natural circulation conditions at the experimental test facility for an integral reactor. A pressure loss should be provided a function of Reynolds number at low velocity conditions in order to simulate the mass flow rate well under natural circulations.

  8. 2015 Annual Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Ponds

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Michael George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

    This report describes conditions and information, as required by the state of Idaho, Department of Environmental Quality Reuse Permit I-161-02, for the Advanced Test Reactor Complex Cold Waste Ponds located at Idaho National Laboratory from November 1, 2014–October 31, 2015. The effective date of Reuse Permit I-161-02 is November 20, 2014 with an expiration date of November 19, 2019.

  9. Risk factors for visible lesions or positive laboratory tests in bovine tuberculosis reactor cattle in Northern Ireland.

    Science.gov (United States)

    O'Hagan, M J H; Courcier, E A; Drewe, J A; Gordon, A W; McNair, J; Abernethy, D A

    2015-07-01

    An observational case-control study was conducted to investigate risk factors for confirmed bovine tuberculosis (bTB) infection in cattle reacting positively to the single intradermal comparative cervical test (SICCT) in Northern Ireland in the years 1998, 2002 and 2006. Macroscopic lesions were detected at slaughter (positive visible lesion (VL) status) in 43.0% of reactor cattle, whilst 45.3% of those sampled were confirmed as bTB positive due to the presence of lesions or positive histopathology/mycobacterial culture (positive bTB status). In 97.5% of the reactors, the VL status and bTB status were either both negative or both positive. Generalized linear mixed model analyses were conducted on data of 24,923 reactor cattle with the variables herd identifier, local veterinary office (DVO) and abattoir being used as random effects within all the models generated at univariable and multivariable level. The other variables within the dataset were used as fixed effects. Significant risk factors associated with VL status and bTB status at multivariable level (pbovine tuberculin injection site, epidemiological status of skin test, total number of reactors at the disclosure test, mean herd size and prior response to the skin test. These risk factors are likely related to the time since infection, the strength of the challenge of infection and the susceptibility of the animal. These findings are important as the detection of visible lesions and the confirmation of bTB are an integral part of the overall bTB control programme in Northern Ireland and the veterinary meat inspection and hygiene programme. The visible lesion status and bTB status of an animal can affect the way in which bTB breakdowns are managed, since failure to detect visible lesions and recovery of Mycobacterium bovis can lead to a less stringent follow-up after other risk factors have been taken into account.

  10. Proceedings of the 4th International Symposium on Material Testing Reactors; December 5-9, 2011, Oarai, Japan

    OpenAIRE

    石原 正博; 鈴木 雅秀

    2012-01-01

    This report is the Proceedings of the 4th International Symposium on Materials Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The 4th symposium was originally scheduled to be held INVAP in Argentina. However, the aftermath of volcanic explosion at Chili forced the symposium to change place. Total 111 participants attended from Argentina, Belgium, France, Germany, Indonesia, Malaysia, Korea, South Africa, Switzerland, the United State and Japan. This symposium addressed the gene...

  11. Deuterium--tritium plasmas in novel regimes in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bell, M.G.; Batha, S.; Beer, M.; Bell, R.E.; Belov, A.; Berk, H.; Bernabei, S.; Bitter, M.; Breizman, B.; Bretz, N.L.; Budny, R.; Bush, C.E.; Callen, J.; Cauffman, S.; Chang, C.S.; Chang, Z.; Cheng, C.Z.; Darrow, D.S.; Dendy, R.O.; Dorland, W.; Duong, H.; Efthimion, P.C.; Ernst, D.; Evenson, H.; Fisch, N.J.; Fisher, R.; Fonck, R.J.; Fredrickson, E.D.; Fu, G.Y.; Furth, H.P.; Gorelenkov, N.N.; Goloborodko, V.Y.; Grek, B.; Grisham, L.R.; Hammett, G.W.; Hawryluk, R.J.; Heidbrink, W.; Herrmann, H.W.; Herrmann, M.C.; Hill, K.W.; Hogan, J.; Hooper, B.; Hosea, J.C.; Houlberg, W.A.; Hughes, M.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Kaita, R.; Kaye, S.; Kesner, J.; Kim, J.S.; Kissick, M.; Krasilnikov, A.V.; Kugel, H.; Kumar, A.; Lam, N.T.; Lamarche, P.; LeBlanc, B.; Levinton, F.M.; Ludescher, C.; Machuzak, J.; Majeski, R.P.; Manickam, J.; Mansfield, D.K.; Mauel, M.; Mazzucato, E.; McChesney, J.; McCune, D.C.; McKee, G.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Mirnov, S.V.; Mueller, D.; Nagayama, Y.; Navratil, G.A.; Nazikian, R.; Okabayashi, M.; Osakabe, M.; Owens, D.K.; Park, H.K.; Park, W.; Paul, S.F.; Petrov, M.P.; Phillips, C.K.; Phillips, M.; Phillips, P.; Ramsey, A.T.; Rice, B.; Redi, M.H.; Rewoldt, G.; Reznik, S.; Roquemore, A.L.; Rogers, J.; Ruskov, E.; Sabbagh, S.A.; Sasao, M.; Schilling, G.; Schmidt, G.L.; Scott, S.D.; Semenov, I.; Senko, T.; Skinner, C.H.; Stevenson, T.; Strait, E.J.; Stratton, B.C.; Strachan, J.D.; Stodiek, W.; Synakowski, E.; Takahashi, H.; Tang, W.; Taylor, G.; Thompson, M.E.; von Goeler, S.; Von Halle, A.; Walters, R.T.; Wang, S.; White, R.; Wieland, R.M.; Williams, M.; Wilson, J.R.; Wong, K.L.; Wurden, G.A.; Yamada, M.; Yavorski, V.; Young, K.M.; Zakharov, L.; Zarnstorff, M.C.; Zweben, S.J. [Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States)

    1997-05-01

    Experiments in the Tokamak Fusion Test Reactor (TFTR) [Phys. Plasmas {bold 2}, 2176 (1995)] have explored several novel regimes of improved tokamak confinement in deuterium{endash}tritium (D--T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high l{sub i}). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through {ital in situ} deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q{sub a}{approx}4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l{sub i} plasmas produced by rapid expansion of the minor cross section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D--T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D--T plasmas with q{sub 0}{gt}1 and weak magnetic shear in the central region, a toroidal Alfvn eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions. {copyright} {ital 1997 American Institute of Physics.}

  12. Deuterium-tritium plasmas in novel regimes in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bell, M.G.; Beer, M. [Princeton Univ., NJ (United States). Princeton Plasma Physics Lab.; Batha, S. [Fusion Physics and Technology, Torrance, CA (United States)] [and others

    1997-02-01

    Experiments in the Tokamak Fusion Test Reactor (TFTR) have explored several novel regimes of improved tokamak confinement in deuterium-tritium (D-T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high-l{sub i}). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in-situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q{sub a} {approx} 4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l{sub i} plasmas produced by rapid expansion of the minor cross-section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D-T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D-T plasmas with q{sub 0} > 1 and weak magnetic shear in the central region, a toroidal Alfven eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode-conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions.

  13. Test problem for thermal-hydraulics and neutronic coupled calculation fore ALFREAD reactor core

    Science.gov (United States)

    Filip, A.; Darie, G.; Saldikov, I. S.; Smirnov, A. D.; Tikhomirov, G. V.

    2017-01-01

    The beginning of a new era of nuclear reactor requires technological advances and also multiples studies. The European Liquid metal cooled Fast breeder Reactor is one of the designs for the generation IV nuclear reactor, selected by ENEA. A pioneer of its time, ELFR needs a demonstrator in order to prove the feasibility of this project and to acquire more data and experience in operating a LFR. For this reason the ALFRED project was started and it is expected to be under operation by the year 2030. This paper has the objective of analyzing the neutronic and thermohydraulics of the ALFRED core by the means of a coupled scheme. The selected code for neutronic simulation is MCNP and the selected code for thermohydraulics is ANSYS.

  14. Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem

    Energy Technology Data Exchange (ETDEWEB)

    Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Grudzinski, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-12-21

    This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.

  15. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

  16. Design of the VISTA-ITL Test Facility for an Integral Type Reactor of SMART and a Post-Test Simulation of a SBLOCA Test

    Directory of Open Access Journals (Sweden)

    Hyun-Sik Park

    2014-01-01

    Full Text Available To validate the performance and safety of an integral type reactor of SMART, a thermal-hydraulic integral effect test facility, VISTA-ITL, is introduced with a discussion of its scientific design characteristics. The VISTA-ITL was used extensively to assess the safety and performance of the SMART design, especially for its passive safety system such as a passive residual heat removal system, and to validate various thermal-hydraulic analysis codes. The VISTA-ITL program includes several tests on the SBLOCA, CLOF, and PRHRS performances to support a verification of the SMART design and contribute to the SMART design licensing by providing proper test data for validating the system analysis codes. A typical scenario of SBLOCA was analyzed using the MARS-KS code to assess the thermal-hydraulic similarity between the SMART design and the VISTA-ITL facility, and a posttest simulation on a SBLOCA test for the shutdown cooling system line break has been performed with the MARS-KS code to assess its simulation capability for the SBLOCA scenario of the SMART design. The SBLOCA scenario in the SMART design was well reproduced using the VISTA-ITL facility, and the measured thermal-hydraulic data were properly simulated with the MARS-KS code.

  17. DEVELOPMENT AND TESTING OF FAULT-DIAGNOSIS ALGORITHMS FOR REACTOR PLANT SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    Grelle, Austin L.; Park, Young S.; Vilim, Richard B.

    2016-06-26

    Argonne National Laboratory is further developing fault diagnosis algorithms for use by the operator of a nuclear plant to aid in improved monitoring of overall plant condition and performance. The objective is better management of plant upsets through more timely, informed decisions on control actions with the ultimate goal of improved plant safety, production, and cost management. Integration of these algorithms with visual aids for operators is taking place through a collaboration under the concept of an operator advisory system. This is a software entity whose purpose is to manage and distill the enormous amount of information an operator must process to understand the plant state, particularly in off-normal situations, and how the state trajectory will unfold in time. The fault diagnosis algorithms were exhaustively tested using computer simulations of twenty different faults introduced into the chemical and volume control system (CVCS) of a pressurized water reactor (PWR). The algorithms are unique in that each new application to a facility requires providing only the piping and instrumentation diagram (PID) and no other plant-specific information; a subject-matter expert is not needed to install and maintain each instance of an application. The testing approach followed accepted procedures for verifying and validating software. It was shown that the code satisfies its functional requirement which is to accept sensor information, identify process variable trends based on this sensor information, and then to return an accurate diagnosis based on chains of rules related to these trends. The validation and verification exercise made use of GPASS, a one-dimensional systems code, for simulating CVCS operation. Plant components were failed and the code generated the resulting plant response. Parametric studies with respect to the severity of the fault, the richness of the plant sensor set, and the accuracy of sensors were performed as part of the validation

  18. Advanced neutron source reactor thermal-hydraulic test loop facility description

    Energy Technology Data Exchange (ETDEWEB)

    Felde, D.K.; Farquharson, G.; Hardy, J.H.; King, J.F.; McFee, M.T.; Montgomery, B.H.; Pawel, R.E.; Power, B.H.; Shourbaji, A.A.; Siman-Tov, M.; Wood, R.J.; Yoder, G.L.

    1994-02-01

    The Thermal-Hydraulic Test Loop (THTL) is a facility for experiments constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory. The ANSR is both cooled and moderated by heavy water and uses uranium silicide fuel. The core is composed of two coaxial fuel-element annuli, each of different diameter. There are 684 parallel aluminum-clad fuel plates (252 in the inner-lower core and 432 in the outer-upper core) arranged in an involute geometry that effectively creates an array of thin rectangular flow channels. Both the fuel plates and the coolant channels are 1.27 mm thick, with a span of 87 mm (lower core), 70 mm (upper core), and 507-mm heated length. The coolant flows vertically upwards at a mass flux of 27 Mg/m{sup 2}s (inlet velocity of 25 m/s) with an inlet temperature of 45{degrees}C and inlet pressure of 3.2 MPa. The average and peak heat fluxes are approximately 6 and 12 MW/m{sup 2}, respectively. The availability of experimental data for both flow excursion (FE) and true critical heat flux (CHF) at the conditions applicable to the ANSR is very limited. The THTL was designed and built to simulate a full-length coolant subchannel of the core, allowing experimental determination of thermal limits under the expected ANSR thermal-hydraulic conditions. For these experimental studies, the involute-shaped fuel plates of the ANSR core with the narrow 1.27-mm flow gap are represented by a narrow rectangular channel. Tests in the THTL will provide both single- and two-phase thermal-hydraulic information. The specific phenomena that are to be examined are (1) single-phase heat-transfer coefficients and friction factors, (2) the point of incipient boiling, (3) nucleate boiling heat-transfer coefficients, (4) two-phase pressure-drop characteristics in the nucleate boiling regime, (5) flow instability limits, and (6) CHF limits.

  19. Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

    Science.gov (United States)

    Briggs, Maxwell H.; Gibson, Marc A.; Sanzi, James

    2017-01-01

    The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one

  20. REMOVAL OF CHEMICAL AND MICROBIAL CONTAMINANTS IN DRINKING WATER - WATTS PREMIER M-2400 POINT-OF-ENTRY REVERSE OSMOSIS DRINKINGWATER TREATMENT SYSTEM

    Science.gov (United States)

    The Watts Premier M-2400 POE RO Drinking Water Treatment System was tested at the NSF Drinking Water Treatment Systems Laboratory for removal of the viruses fr and MS2, the bacteria Brevundimonas diminuta, and chemicals aldicarb, benzene, cadmium, carbofuran, cesium, chl...

  1. REMOVAL OF CHEMICAL AND MICROBIAL CONTAMINANTS IN DRINKING WATER - WATTS PREMIER M-2400 POINT-OF-ENTRY REVERSE OSMOSIS DRINKINGWATER TREATMENT SYSTEM

    Science.gov (United States)

    The Watts Premier M-2400 POE RO Drinking Water Treatment System was tested at the NSF Drinking Water Treatment Systems Laboratory for removal of the viruses fr and MS2, the bacteria Brevundimonas diminuta, and chemicals aldicarb, benzene, cadmium, carbofuran, cesium, chl...

  2. Status of Wrought FeCrAl-UO2 Capsules Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harp, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Core, G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-07-01

    Candidate cladding materials for accident tolerant fuel applications require extensive testing and validation prior to commercial deployment within the nuclear power industry. One class of cladding materials, FeCrAl alloys, is currently undergoing such effort. Within these activities is a series of irradiation programs within the Advanced Test Reactor. These programs are developed to aid in commercial maturation and understand the fundamental mechanisms controlling the cladding performance during normal operation of a typical light water reactor. Three different irradiation programs are on-going; one designed as a simple proof-of-principle concept, the other to evaluate the susceptibility of FeCrAl to fuel-cladding chemical interaction, and the last to fully simulate the conditions of a pressurized water reactor experimentally. To date, nondestructive post-irradiation examination has been completed on the rodlet deemed FCA-L3 from the simple proof-of-concept irradiation program. Initial results show possible breach of the rodlet under irradiation but further studies are needed to conclusively determine whether breach has occurred and the underlying reasons for such a possible failure. Further work includes characterizing additional rodlets following irradiation.

  3. Methodology comparison for gamma-heating calculations in material-testing reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lemaire, M.; Vaglio-Gaudard, C.; Lyoussi, A. [CEA, DEN, DER, Cadarache F-13108 Saint Paul les Durance (France); Reynard-Carette, C. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France)

    2015-07-01

    The Jules Horowitz Reactor (JHR) is a Material-Testing Reactor (MTR) under construction in the south of France at CEA Cadarache (French Alternative Energies and Atomic Energy Commission). It will typically host about 20 simultaneous irradiation experiments in the core and in the beryllium reflector. These experiments will help us better understand the complex phenomena occurring during the accelerated ageing of materials and the irradiation of nuclear fuels. Gamma heating, i.e. photon energy deposition, is mainly responsible for temperature rise in non-fuelled zones of nuclear reactors, including JHR internal structures and irradiation devices. As temperature is a key parameter for physical models describing the behavior of material, accurate control of temperature, and hence gamma heating, is required in irradiation devices and samples in order to perform an advanced suitable analysis of future experimental results. From a broader point of view, JHR global attractiveness as a MTR depends on its ability to monitor experimental parameters with high accuracy, including gamma heating. Strict control of temperature levels is also necessary in terms of safety. As JHR structures are warmed up by gamma heating, they must be appropriately cooled down to prevent creep deformation or melting. Cooling-power sizing is based on calculated levels of gamma heating in the JHR. Due to these safety concerns, accurate calculation of gamma heating with well-controlled bias and associated uncertainty as low as possible is all the more important. There are two main kinds of calculation bias: bias coming from nuclear data on the one hand and bias coming from physical approximations assumed by computer codes and by general calculation route on the other hand. The former must be determined by comparison between calculation and experimental data; the latter by calculation comparisons between codes and between methodologies. In this presentation, we focus on this latter kind of bias. Nuclear

  4. Radiation-induced electrical degradation experiments in the Japan materials testing reactor

    Science.gov (United States)

    Farnum, Eugene H.; Shikama, Tatsuo; Narui, Minoru; Sagawa, Tsutomu; Scarborough, Kent

    1996-02-01

    An experiment to measure radiation-induced electrical degradation (RIED) in a sapphire sample and in three MgO-insulated cables was conducted at the JMTR light water reactor. The materials were irradiated at about 260°C to a fluence of 3 × 1024 n/m 2 ( E > 1 MeV) with an applied DC electric field between 100 kV/m and 500 kV/m. Even though the results for the sapphire sample are somewhat ambiguous because of an unexplained offset current of about 0.6 μA substantial degradation was not observed in the sapphire: instead, radiation-induced conductivity (RIC) seemed to decrease slightly during the experiment. Substantial increase in leakage current, that increased with applied electric field, occurred in the MgO-insulated cables. This increased conductivity disappeared when the reactor was shut down and sample temperature returned to ambient. However, the physical degradation apparently remained in the material while the reactor was off because restarting the irradiation brought the conductivity back to its previous, degraded, reactor-on value. This effect is different from the RIED effect reported by Hodgson but is similar to previous results reported by Shikama et al. Considerable data were taken to determine the sample temperature and leakage currents during the irradiation.

  5. Tokamak Fusion Test Reactor. Final conceptual design report. [Overall cost and scheduling program

    Energy Technology Data Exchange (ETDEWEB)

    1976-02-01

    The TFTR is the first U.S. magnetic confinement device planned to demonstrate the fusion of D-T at reactor power levels. This report addresses the physics objectives and the engineering goals of the TFTR project. Technical, cost, and schedule aspects of the project are included. (MOW)

  6. Emission and transmission tomography systems to be developed for the future needs of Jules Horowitz material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kotiluoto, Petri [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland)], E-mail: petri.kotiluoto@vtt.fi; Wasastjerna, Frej; Kekki, Tommi [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland); Sipilae, Heikki; Banzuzi, Kukka [Oxford Instruments Analytical Oy, Nihtisillankuja 5, P.O.Box 85, FI-02631 Espoo (Finland); Kinnunen, Petri; Heikinheimo, Liisa [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland)

    2009-08-01

    The new 100 MW Jules Horowitz material testing reactor will be built in Cadarache, France. It will support, for instance, research on new types of innovative nuclear fuel. As a Finnish in-kind contribution, 3D emission and transmission tomography equipment will be delivered for both the reactor and the active component storage pool. The image reconstruction of activities inside the used nuclear fuel will be based on gamma spectrometry measurements. A new type of underwater digital X-ray linear detector array is under development for transmission imaging, based on GaAs and direct conversion of X-rays into an electrical signal. A shared collimator will be used for both emission and transmission measurements. Some preliminary design has been performed. For the current design, the expected gamma spectrometric response of a typical high-purity germanium detector has been simulated with MCNP for minimum and maximum source activities (specified by CEA) to be measured in future.

  7. Cumulative damage fatigue tests on nuclear reactor Zircaloy-2 fuel tubes at room temperature and 300°C

    Science.gov (United States)

    Pandarinathan, P. R.; Vasudevan, P.

    1980-06-01

    Cumulative damage fatigue tests were conducted on the Zircaloy-2 fuel tubes at room temperature and 300°C on the modified Moore type, four-point-loaded, deflection-controlled, rotating bending fatigue testing machine. The cumulative cycle ratio at fracture for the Zircaloy-2 fuel tubes was found to depend on the sequence of loading, stress history, number of cycles of application of the pre-stress and the test temperature. A Hi-Lo type fatigue loading was found to be very much damaging at room temperature and this feature was not observed in the tests at 300°C. Results indicate significant differences in damage interaction and damage propagation under cumulative damage tests at room temperature and at 300°C. Block-loading fatigue tests are suggested as the best method to determine the life-time of Zircaloy-2 fuel tubes under random fatigue loading during their service in the reactor.

  8. Visible-Light-Responsive Photocatalysis: Ag-Doped TiO2 Catalyst Development and Reactor Design Testing

    Science.gov (United States)

    Coutts, Janelle L.; Hintze, Paul E.; Meier, Anne; Shah, Malay G.; Devor, Robert W.; Surma, Jan M.; Maloney, Phillip R.; Bauer, Brint M.; Mazyck, David W.

    2016-01-01

    In recent years, the alteration of titanium dioxide to become visible-light-responsive (VLR) has been a major focus in the field of photocatalysis. Currently, bare titanium dioxide requires ultraviolet light for activation due to its band gap energy of 3.2 eV. Hg-vapor fluorescent light sources are used in photocatalytic oxidation (PCO) reactors to provide adequate levels of ultraviolet light for catalyst activation; these mercury-containing lamps, however, hinder the use of this PCO technology in a spaceflight environment due to concerns over crew Hg exposure. VLR-TiO2 would allow for use of ambient visible solar radiation or highly efficient visible wavelength LEDs, both of which would make PCO approaches more efficient, flexible, economical, and safe. Over the past three years, Kennedy Space Center has developed a VLR Ag-doped TiO2 catalyst with a band gap of 2.72 eV and promising photocatalytic activity. Catalyst immobilization techniques, including incorporation of the catalyst into a sorbent material, were examined. Extensive modeling of a reactor test bed mimicking air duct work with throughput similar to that seen on the International Space Station was completed to determine optimal reactor design. A bench-scale reactor with the novel catalyst and high-efficiency blue LEDs was challenged with several common volatile organic compounds (VOCs) found in ISS cabin air to evaluate the system's ability to perform high-throughput trace contaminant removal. The ultimate goal for this testing was to determine if the unit would be useful in pre-heat exchanger operations to lessen condensed VOCs in recovered water thus lowering the burden of VOC removal for water purification systems.

  9. Tears of a Chinese Premier Hussein Ismail Hussein

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    THE day Premier Wen took office he stated.""Leaders should be closer to the masses."" His visit to comfort and talk with everyday workers in Tongchuan City, Shaanxi Province on January 1, 2005, where a few weeks previously a gas blast in the Chenjiashan Coalmine had killed 166 workers, was by no means his first. But it was the first time he publicly shed tears. In one household that had lost its breadwinner, Wen Jiabao embraced the victim's son and shared his grief. He later had a simple lunch of steamed bread and tea in a tunnel 1,300 meters belowg round as he chatted with workers at another mine in the city.

  10. L’abri premier, de Vitruve à Nils-Udo

    Directory of Open Access Journals (Sweden)

    Nathalie Huvenne

    2011-05-01

    Full Text Available Cette étude se propose de mettre en écho les écrits de Vitruve, livre second, chapitre 1 du De architectura et les créations plastiques de l’artiste contemporain Nils-Udo.Ce vis-à-vis des réflexions de Vitruve avec celles de l’artiste nous permettra de tisser un lien entre ces deux hommes et de voir dans leurs travaux respectifs des points de similitude.Le texte extrait du De architectura II, 1, sous-titré par M. Nisard, De la manière de vivre des premiers hommes, et quels ont été les commenc...

  11. Component test for validation of the design life of reactor vessel wall of the fast breeder reactor SNR 300 regarding creep-fatigue. Modellversuch zur Absicherung der Auslegung der Reaktortankwand des SNR 300 hinsichtlich Kriechermuedung

    Energy Technology Data Exchange (ETDEWEB)

    Lohse, U. (Interatom GmbH, Bergisch Gladbach (Germany)); Laue, H. (Interatom GmbH, Bergisch Gladbach (Germany)); Rathjen, P. (Interatom GmbH, Bergisch Gladbach (Germany)); Maile, K. (Staatliche Materialpruefungsanstalt, Stuttgart (Germany)); Eckert, W. (Staatliche Materialpruefungsanstalt, Stuttgart (Germany)); Purper, H. (Staatliche Materialpruefungsanstalt, Stuttgart (Germany))

    1991-01-01

    The design of the reactor vessel wall is tested under a long-term creep-fatigue stress, with the aid of a similar components test. For this purpose, the results of calculation are compared with the experimental results concerning deformation depending on point and time, and damage depending on the initial state of material, point and time. (orig./HP)

  12. Premier Li Keqiang and Indian Prime Minister Modi Attend Regional Forum

    Institute of Scientific and Technical Information of China (English)

    Zhang; Min

    2015-01-01

    The First Forum of Leaders of the Regions of China and India,cosponsored by the CPAFFC and the China International Friendship Cities Association(CIFCA),was held at the Great Hall of the People in Beijing on May 15.Chinese Premier Li Keqiang and his Indian counterpart Narendra Modi gave addresses.Premier Li expressed his congrat-

  13. Standard Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance, E706(IIIB)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This test method describes the use of solid-state track recorders (SSTRs) for neutron dosimetry in light-water reactor (LWR) applications. These applications extend from low neutron fluence to high neutron fluence, including high power pressure vessel surveillance and test reactor irradiations as well as low power benchmark field measurement. (1) This test method replaces Method E 418. This test method is more detailed and special attention is given to the use of state-of-the-art manual and automated track counting methods to attain high absolute accuracies. In-situ dosimetry in actual high fluence-high temperature LWR applications is emphasized. 1.2 This test method includes SSTR analysis by both manual and automated methods. To attain a desired accuracy, the track scanning method selected places limits on the allowable track density. Typically good results are obtained in the range of 5 to 800 000 tracks/cm2 and accurate results at higher track densities have been demonstrated for some cases. (2) Trac...

  14. Premier`s Task Force on NAFTA wind turbine manufacturing facility and windpower plants : final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The proposal for a 700 MW wind farm and associated manufacturing facility at Pincher Creek, Alberta was discussed. York WindPower and ENERCON submitted the joint proposal in the spring of 1996 and requested a financial arrangement to guarantee a sale price of 5.4 cents per kWh (escalated) over 25 years. This was later revised in February, 1997 to 350 MW, 4.9 cents per kWh (escalated) over 15 years. A Task Force was established to assess this proposal and any other prospects for development of renewables in general. The two inseparable elements of the proposed project would be a manufacturing facility which would produce approximately 400 wind turbines a year, and a 700 MW wind farm, phased in over 10 years. The size of the wind farm would be based on a calculation of the minimal annual production required for the manufacturing facility to be viable. Pincher Creek residents are supportive of renewable energy and have been promoting wind energy long before the York/ENERCON proposal. They view this project as a support for regional economic development. The Task Force was pleased that York/ENERCON is pursuing the Alberta Advantage and is considering setting up a manufacturing facility. The Task Force did not seek to make a finding with respect to the business viability of the project, instead, the Task Force concluded that the commercial test for the project should be provided by competition in the marketplace. Nevertheless, the Task Force is supportive of the project, provided it proceeds on a voluntary, market-driven basis, and there are no conflicts with the existing government policy framework. 4 tabs., 2 appendices.

  15. Speech by Prime Minister Francois Fillon. Visit of the Jules Horowitz experimental reactor works on the Commissariat a l'Energie et aux Energies Alternatives site. Cadarache, May 3, 2010; Discours du Premier ministre Francois FILLON Cadarache, lundi 3 mai 2010. Visite du chantier du Reacteur experimental Jules Horowitz sur le site du Commissariat a l'Energie Atomique et aux Energies Alternatives

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-07-01

    In this speech, the French Prime Minister evokes the present context, the importance of strategic technologies, and the challenge of investing in these technologies within a context of reduction of public expenses. He comments the decision of his government to finance research and education activities in different domains, and more specifically in the energy sector with this fourth generation Jules Horowitz experimental reactor. He recalls that the nuclear sector has always been very important to the eyes of the successive French governments, and outlines how this reactor will contribute to reactor operational optimization, lifetime extension and safety, nuclear fuel development, etc.

  16. RJH, a new test reactor in Europe; Le RJH - un nouveau reacteur d'essai en europe

    Energy Technology Data Exchange (ETDEWEB)

    Iracane, D. [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), 91 - Gif sur Yvette (France)

    2005-07-01

    Material test reactors (MTR) are now ageing in Europe and they cannot secure the experimental needs over next decades. In this context, a new MTR, named Jules Horowitz reactor (RJH), operated as an international user-facility, is under development on the Cea's site of Cadarache (France). The design studies will end in 2007, the construction stage will follow and RJH commissioning is scheduled in 2014. Its construction costs are estimated to 500 million euros. RJH is a pool reactor of 100 MWth, its core will be inserted in a pressurized vessel with a primary circuit assuring water flow through forced convection. The core inlet-outlet temperature is about 25-40 Celsius degrees. RJH core is designed to use a high density - low enrichment UMo nuclear fuel (8 gU/cm{sup 3}, enrichment rate: 19.75%). Experimental devices located in the core will benefit from neutron fluxes ranging from 2.5 10{sup 14} n/cm{sup 2}.s to 5.10{sup 14} n/cm{sup 2}.s (E > 1 MeV). RJH is designed to manage simultaneously 10 experiments in the core and as many in the reflector. (A.C.)

  17. Thermal tests of a multi-tubular reactor for hydrogen production by using mixed ferrites thermochemical cycle

    Science.gov (United States)

    Gonzalez-Pardo, Aurelio; Denk, Thorsten; Vidal, Alfonso

    2017-06-01

    The SolH2 project is an INNPACTO initiative of the Spanish Ministry of Economy and Competitiveness, with the main goal to demonstrate the technological feasibility of solar thermochemical water splitting cycles as one of the most promising options to produce H2 from renewable sources in an emission-free way. A multi-tubular solar reactor was designed and build to evaluate a ferrite thermochemical cycle. At the end of this project, the ownership of this plant was transferred to CIEMAT. This paper reviews some additional tests with this pilot plant performed in the Plataforma Solar de Almería with the main goal to assess the thermal behavior of the reactor, evaluating the evolution of the temperatures inside the cavity and the relation between supplied power and reached temperatures. Previous experience with alumina tubes showed that they are very sensitive to temperature and flux gradients, what leads to elaborate an aiming strategy for the heliostat field to achieve a uniform distribution of the radiation inside the cavity. Additionally, the passing of clouds is a phenomenon that importantly affects all the CSP facilities by reducing their efficiency. The behavior of the reactor under these conditions has been studied.

  18. Irradiation Test of RPV Steel in 49-2 Research Reactor

    Institute of Scientific and Technical Information of China (English)

    LIN; Yun; TONG; Zhen-feng; ZHANG; Chang-yi; NING; Guang-sheng; YANG; Wen

    2013-01-01

    Reactor pressure vessel(RPV)is the critical and unchangeable component of the PWR during its service lifetime which determines the lifetime of nuclear power plant(NPP).The property decline of RPVis mainly caused by large dose and high energy neutron irradiation during the long term service.Irradiation can reduce the toughness and increase the brittleness of the RPV materials.Thus raising the

  19. Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

    2007-10-01

    The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

  20. Evaluation of ductile-brittle transition behavior with neutron irradiation in nuclear reactor pressure vessel steels using small punch test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. C.; Lee, B. S. [KAERI, Taejon (Korea, Republic of); Oh, Y. J. [Hanbat National Univ., Taejon (Korea, Republic of)

    2003-10-01

    A Small Punch (SP) test was performed to evaluate the ductile-brittle transition temperature before and after neutron irradiation in Reactor Pressure Vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the Charpy test and Master Curve fracture toughness test in accordance with the ASTM standard E1921. The samples were taken from 1/4t location of the vessel thickness and machined into a 10x10x0.5mm dimension. Irradiation of the samples was carried out in the research reactor at KAERI (HANARO) at about 290 .deg. C of the different fluence levels respectively. SP tests were performed in the temperature range of RT to -196 .deg. C using a 2.4mm diameter ball. For the materials before and after irradiation, SP transition temperatures (T{sub sp}), which are determined at the middle of the upper and lower SP energies, showed a linear correlation with the Charpy index temperature, T{sub 41J}. T{sub sp} from the irradiated samples was increased as the fluence level increased and was well within the deviation range of the unirradiated data. The TSP had a correlation with the reference temperature (T{sub 0}) from the master curve method using a pre-cracked Charpy V-notched (PCVN) specimen.

  1. Design and Status of the NGNP Fuel Experiment AGR-3/4 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in November 2013. Since the purpose of this experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  2. Development and testing of a recorder and controller for a microalgae culture reactor; Desarrollo y prueba de un registrador y controlador para un reactor de cultivo de microalgas

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel, Wilson; Reyes, Jose Fernando; Bruijn, Johannes; Hernandez, Alejandro [Universidad de Concepcion, Chilan (Chile). Facultad de Ingenieria Agricola. Dept. de Mecanizacion y Energia], Emails: wesquive@udec.cl., jreyes@udec.cl., jdebruij@udec.cl., alehernandez@udec.cl

    2010-07-01

    An electronic system to monitor and control operational variables in a Raceway type of reactor for the culture of the Scenedesmus spinosus microalgae and later production for biodiesel and mitigating CO{sub 2} was developed and tested. The electronic system is constituted by a micro controller, a card reader SD, a card SD, a real-time clock, a power supply, a screen GLCD, a keyboard and a card for data acquisition, all implemented for 4-20 mA and 0-5 V output sensors. Temperature, pH, electrical conductivity, dissolved oxygen and solar radiation were measured digitalized and saved every 10 minutes. These variables were digitalized and kept in the SD memory every 10 minutes. It was determined that the most favorable conditions for the proliferation of the culture are near pH neutral and a temperature of 30 deg C, existing a strong correlation between pH and the dissolved CO{sub 2} level. Using the digital outputs of temperature and pH of the microcontroller, the CO{sub 2} injection and the elimination of O{sub 2} were controlled to maintain an adequate environment for the development of the culture. (author)

  3. In-Pile Experiment of a New Hafnium Aluminide Composite Material to Enable Fast Neutron Testing in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Douglas L. Porter; James R. Parry; Heng Ban

    2010-06-01

    A new hafnium aluminide composite material is being developed as a key component in a Boosted Fast Flux Loop (BFFL) system designed to provide fast neutron flux test capability in the Advanced Test Reactor. An absorber block comprised of hafnium aluminide (Al3Hf) particles (~23% by volume) dispersed in an aluminum matrix can absorb thermal neutrons and transfer heat from the experiment to pressurized water cooling channels. However, the thermophysical properties, such as thermal conductivity, of this material and the effect of irradiation are not known. This paper describes the design of an in-pile experiment to obtain such data to enable design and optimization of the BFFL neutron filter.

  4. The reactor Cabri; La pile cabri

    Energy Technology Data Exchange (ETDEWEB)

    Ailloud, J.; Millot, J.P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1964-07-01

    It has become necessary to construct in France a reactor which would permit the investigation of the conditions of functioning of future installations, the choice, the testing and the development of safety devices to be adopted. A water reactor of a type corresponding to the latest CEA constructions in the field of laboratory or university reactors was decided upon: it appeared important to be able to evaluate the risks entailed and to study the possibilities of increasing the power, always demanded by the users; on the other hand, it is particularly interesting to clarify the phenomena of power oscillation and the risks of burn out. The work programme for CABRI will be associated with the work carried out on the American Sperts of the same type, during its construction, very useful contacts were made with the American specialists who designed the se reactors. A brief description of the reactor is given in the communication as well as the work programme for the first years with respect to the objectives up to now envisaged. Rough description of the reactor. CABRI is an open core swimming-pool reactor without any lateral protection, housed in a reinforced building with controlled leakage, in the Centre d'Etudes Nucleaires de Cadarache. It lies alone in the middle of an area whose radius is 300 meters long. Control and measurements equipment stand out on the edge of that zone. It consumes MTR fuel elements. The control-safety rods are propelled by compressed air. The maximum flow rate of cooling circuit is 1500 m{sup 3}/h. Transient measurements are recorded in a RW330 unit. Aims and work programme. CABRI is meant for: - studies on the safety of water reactors - for the definition of the safety margins under working conditions: research of maximum power at which a swimming-pool reactor may operate with respect to a cooling accident, of local boiling effect on the nuclear behaviour of the reactor, performances of the control and safety instruments under

  5. Leakage Tests of the Stainless Steel Vessels of the Antineutrino Detectors in the Daya Bay Reactor Neutrino Experiment

    CERN Document Server

    Chen, Xiaohui; Heng, Yuekun; Wang, Lingshu; Tang, Xiao; Ma, Xiaoyan; Zhuang, Honglin; Band, Henry; Cherwinka, Jeff; Xiao, Qiang; Heeger, Karsten M

    2012-01-01

    The antineutrino detectors in the Daya Bay reactor neutrino experiment are liquid scintillator detectors designed to detect low energy particles from antineutrino interactions with high efficiency and low backgrounds. Since the antineutrino detector will be installed in a water Cherenkov cosmic ray veto detector and will run for 3 to 5 years, ensuring water tightness is critical to the successful operation of the antineutrino detectors. We choose a special method to seal the detector. Three leak checking methods have been employed to ensure the seal quality. This paper will describe the sealing method and leak testing results.

  6. On0Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ayman I. Hawari; Mohamed A. Bourham

    2010-04-22

    IVery High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (~ 1- mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4% – 10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

  7. [Radiation ecological environment in the Republic of Kazakhstan in the vicinity of the reactors and on the territory of the Semipalatinsk Test Site].

    Science.gov (United States)

    Kim, D S

    2012-01-01

    The results of research into the environmental conditions in the regions of location of the pressurized water reactor WWR-K, fast neutron breeder BN-350 and on the territory of the Semipalatinsk Test Site are represented. The effects of the exposure to aerosol emissions from WWR-K and BN-350 reactors on the environment are summarized. We present some arguments in favor of the safe operation of fission reactors in compliance with the rules and norms of nuclear and radiation protection and the efficient disposal of radioactive waste on the territory of the Republic.

  8. Technology, safety, and costs of decommissioning reference nuclear research and test reactors: sensitivity of decommissioning radiation exposure and costs to selected parameters

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.

    1983-07-01

    Additional analyses of decommissioning at the reference research and test (R and T) reactors and analyses of five recent reactor decommissionings are made that examine some parameters not covered in the initial study report (NUREG/CR-1756). The parameters examined for decommissioning are: (1) the effect on costs and radiation exposure of plant size and/or type; (2) the effects on costs of increasing disposal charges and of unavailability of waste disposal capacity at licensed waste disposal facilities; and (3) the costs of and the available alternatives for the disposal of nuclear R and T reactor fuel assemblies.

  9. DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-TEMPERATURE GAS-COOLED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James; Bayless, Paul; Strydom, Gerhard; Kumar, Akansha; Gougar, Hans

    2016-11-01

    Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations. Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.

  10. Radiation tests of ITER diagnostic system materials in the BOR-60 reactor

    Science.gov (United States)

    Revyakin, Yu. L.; Kosenkov, V. M.; Bender, S. E.; Belyakov, V. A.

    1996-10-01

    An in-pile experiment was conducted up to a fluence of 9.9 × 10 25 m 2 ( E ≫ 0.1 MeV) which investigated the following electrophysical characteristics of a cable with mineral insulation and nickel conductor: insulation resistance, radiation-induced current and EMF. Irradiation was also performed in the BOR-60 reactor up to a fluence 10 23 m -2 on six crystal types for the monochromator: mica, LiF, multilayered mirrors Fe/C, W/Si, Cr/C and Mo/Si. Change of the reflectivity, width and shape of diffraction reflections were investigated.

  11. Halden In-Reactor Test to Exhibit PWR Axial Offset Anomaly

    Energy Technology Data Exchange (ETDEWEB)

    P.Bennett, B. Beverskog, R.Suther

    2004-12-01

    Many PWRs have encountered the axial offset anomaly (AOA) since the early 1990s, and these experiences have been reported widely. AOA is a phenomenon associated with localized boron hideout in corrosion product deposits (crud) on fuel surfaces. Several mitigation approaches have been developed or are underway to either delay the onset of AOA or avoid it entirely. This study describes the first phase of an experimental program designed to investigate whether the use of enriched boric acid (EBA) in the reactor coolant can mitigate AOA.

  12. Direct test of the time-independence of fundamental nuclear constants using the Oklo natural reactor

    CERN Document Server

    Shlyakhter, A I

    The positions of neutron resonances have been shown to be highly sensitive to the variation of fundamental nuclear constants. The analysis of the measured isotopic shifts in the natural fossil reactor at Oklo gives the following restrictions on the possible rates of the interaction constants variation: strong ~2x10^-19 yr^-1, electromagnetic ~5x10^-18 yr^-1, weak ~10^-12 yr^-1. These limits permit to exclude all the versions of nuclear constants contemporary variation discussed in the literature. URL: http://alexonline.info >. For more recent analyses see hep-ph/9606486, hep-ph/0205206 and astro-ph/0204069 .

  13. Protective structures on the surface of zirconium components of light water reactor cores: Formation, testing, and prototype equipment

    Energy Technology Data Exchange (ETDEWEB)

    Begrambekov, L. B.; Gordeev, A. A.; Evsin, A. E., E-mail: evsin@plasma.mephi.ru; Ivanova, S. V.; Kaplevsky, A. S.; Sadovskiy, Ya. A. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)

    2015-12-15

    The results of tests of plasma treatment of zirconium and deposition of protective yttrium coatings used as the methods of protection of zirconium components of light water reactor cores against hydrogenation are detailed. The amount of hydrogen in the treated sample exposed to superheated steam for 2500 h at temperature T = 400°C and pressure p = 1 atm was five times lower than the corresponding value for the untreated one. The amount of hydrogen in the sample coated with yttrium remained almost unchanged in 4000 h of exposure. A plasma method for rapid testing for hydrogen resistance is proposed. The hydrogenation rate provided by this method is 700 times higher than that in tests with superheated steam. The results of preliminary experiments confirm the possibility of constructing a unit for batch processing of the surfaces of fuel rod claddings.

  14. Effective Factors on Reducing the Number of Spectators in Iran Football Premier League

    Directory of Open Access Journals (Sweden)

    Amir Reza Khadem Azghadi

    2016-07-01

    Full Text Available Because of reducing the number of spectators of football premier league, this study is seeking for identifying factors put the most impact on this decline. The statistical population consisted of all spectators in Iran football premier league in 2015-16, out of which 395 spectators were randomly selected as the research samples. The data were collected via a researcher-made questionnaire. The first part of the questionnaire included demographic information and the second part, at 6 aspects, includes 35 questions analyzing the reasons for reducing the number of spectators in Iran's football Premier League. For analyzing data, it was used from first and second order confirmatory factor analysis based on structural equations through using SPSS 20 and LISREL 8.8 software. The results of first order confirmatory factor analysis showed that the measurement model of factors affecting on reducing the number of spectators of football premier league is an appropriate model and model parameters are significant. All factors are approved as effective variables on reducing the number of spectators of football premier league. The results also showed that the second order measurement model of effective factors on reducing the number of spectators of football premier league are also appropriate, and economic, facilitative, administrative, technical, cultural-social, and personal-family respectively put the most effects on reducing the number of spectators of football premier league. It is suggested for the sport marketers to analyze identified factors in this research and develop applicable strategies and guidelines for them.

  15. Recent view to the results of pulse tests in the IGR reactor with high burn-up fuel

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L. [Russian Research Centre, Moscow (Russian Federation)

    1996-03-01

    Testing of 43 fuel elements (13 fuel elements with high burn-up fuel, 10 fuel elements with preirradiated cladding and fresh fuel, and 20 non-irradiated fuel elements) was carried out in the IGR pulse reactor with a half width of the reactor power pulse of about 0.7 sec. Tests were conducted in capsules with no coolant flow and with standard initial conditions in the capsule of 20{degrees}C and 0.2 MPa. Two types of coolant were used: water and air. One purpose of the test program was to determine the thresholds and mechanisms of fuel rod failure under RIA conditions for VVER fuel rods over their entire exposure range, from zero to high burn-up. These failure thresholds are often used in safety analyses. The tests and analyses were designed to reveal the influence on fuel rod failure of (1) the mechanical properties of the cladding, (2) the pellet-to-cladding gap, (3) fuel burn-up, (4) fuel-to-coolant heat transfer, and other parameters. The resulting data base can also be used for validation of computer codes used for analyzing fuel rod behavior. Three types of test specimens were used in the tests, and diagrams of these specimens are shown in Fig. 1. {open_quotes}Type-C{close_quotes} specimens were re-fabricated from commercial fuel rods of the VVER-1000 type that had been subjected to many power cycles of operation in the Novovoronezh Nuclear Power Plant (NV NPP). {open_quotes}Type-D{close_quotes} specimens were fabricated from the same commercial fuel rods used above, but the high burn-up oxide fuel was removed from the cladding and was replaced with fresh oxide fuel pellets. {open_quotes}Type-D{close_quotes} specimens thus provided a means of separating the effects of the cladding and the oxide fuel pellets and were used to examine cladding effects only.

  16. Pilot-scale tests of an innovative 'serial self-turning reactor' composting technology in Thailand.

    Science.gov (United States)

    Sungsomboon, Praj-ya; Chaisomphob, Taweep; Bongochgetsakul, Nattakorn; Ishida, Tetsuya

    2013-02-01

    Composting facilities in Thailand have faced various operational problems, resulting in the emission of odours, incomplete digestion of waste organics, and higher than desired costs. Composting technologies imported from developed countries tend to be sized for larger communities and are otherwise not suited for the rural communities that comprise more than 80% of all communities in Thailand. This article addresses the research and development of a novel composting technology aimed at filling this observed need. The study was divided into two parts: (1) the development of a new composting technology and fabrication of a prototype configuration of equipment; and (2) scale-up and study on a pilot-scale using real rubbish. The proposed technology, called 'serial self-turning reactor (STR)', entailed a vertical flow composting system that consisted of a set of aerobic reactors stacked on a set of gravity fed turning units. In-vessel bioreactor technology enables the operator to control composting conditions. The researchers found that a tower-like STR results in flexibility in size scale and waste processing residence time. The pilot-scale experiments showed that the proposed system can produce good quality compost while consuming comparatively little energy and occupying a compact space, compared to traditional land-intensive windrow composting facilities.

  17. KüFA safety testing of HTR fuel pebbles irradiated in the High Flux Reactor in Petten

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O., E-mail: oliver.seeger@rwth-aachen.de [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Laurie, M., E-mail: mathias.laurie@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Abjani, A. El; Ejton, J.; Boudaud, D.; Freis, D.; Carbol, P.; Rondinella, V.V. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Fütterer, M. [European Commission, Joint Research Centre (JRC), Institute for Energy and Transport (IET), Nuclear Reactor Integrity Assessment and Knowledge Management Unit, PO Box 2, 1755 ZG Petten (Netherlands); Allelein, H.-J. [Lehrstuhl für Reaktorsicherheit und -technik an der RWTH Aachen, Kackertstraße 9, 52072 Aachen (Germany)

    2016-09-15

    The Cold Finger Apparatus (KühlFinger-Apparatur—KüFA) in operation at JRC-ITU is designed to experimentally scrutinize the effects of Depressurization LOss of Forced Circulation (D-LOFC) accident scenarios on irradiated High Temperature Reactor (HTR) fuel pebbles. Up to 1600 °C, the reference maximum temperature for these accidents, high-quality German HTR fuel pebbles have already demonstrated a small fission product release. This paper discusses and compares the releases obtained from KüFA-testing the pebbles HFR-K5/3 and HFR-EU1/3, which were both irradiated in the High Flux Reactor (HFR) in Petten. We present the time-dependent fractional release of the volatile fission product {sup 137}Cs as well as the fission gas {sup 85}Kr for both pebbles. For HFR-EU1/3 the isotopes {sup 134}Cs and {sup 154}Eu as well as the shorter-lived {sup 110m}Ag have also been measured. A detailed description of the experimental setup and its accuracy is given. The data for the recently tested pebbles is discussed in the context of previous results.

  18. Flow Components in a NaK Test Loop Designed to Simulate Conditions in a Nuclear Surface Power Reactor

    Science.gov (United States)

    Polzin, Kurt A.; Godfroy, Thomas J.

    2008-01-01

    A test loop using NaK as the working fluid is presently in use to study material compatibility effects on various components that comprise a possible nuclear reactor design for use on the lunar surface. A DC electromagnetic (EM) pump has been designed and implemented as a means of actively controlling the NaK flow rate through the system and an EM flow sensor is employed to monitor the developed flow rate. These components allow for the matching of the flow rate conditions in test loops with those that would be found in a full-scale surface-power reactor. The design and operating characteristics of the EM pump and flow sensor are presented. In the EM pump, current is applied to a set of electrodes to produce a Lorentz body force in the fluid. A measurement of the induced voltage (back-EMF) in the flow sensor provides the means of monitoring flow rate. Both components are compact, employing high magnetic field strength neodymium magnets thermally coupled to a water-cooled housing. A vacuum gap limits the heat transferred from the high temperature NaK tube to the magnets and a magnetically-permeable material completes the magnetic circuit. The pump is designed to produce a pressure rise of 5 psi, and the flow sensor's predicted output is roughly 20 mV at the loop's nominal flow rate of 0.5 GPM.

  19. Cold test with a benchtop set-up for fluidized bed reactor using quartz sand to simulate gasification of coal cokes by concentrated solar radiation

    Science.gov (United States)

    Gokon, Nobuyuki; Tanabe, Tomoaki; Shimizu, Tadaaki; Kodama, Tatsuya

    2016-05-01

    The impacts of internal circulation of a mixture of coal-coke particles and quartz sand on the fluidization state in a fluidized bed reactor are investigated by a cold test with a benchtop set-up in order to design 10-30 kWth scale prototype windowed fluidized-bed reactor. Firstly, a basic relationship between pressure loss of inlet gas and gas velocity was experimentally examined using quartz sand with different particle sizes by a small-scale quartz tube with a distributor at ambient pressure and temperature. Based on the results, an appropriate particle range of quartz sand and layer height/layer diameter ratio (L/D ratio) was determined for a design of the fluidized bed reactor. Secondly, a windowed reactor mock-up was designed and fabricated for solar coke gasification using quartz sand as a bed material. The pressure loss between the inlet and outlet gases was examined, and descending cokes and sand particles on the sidewall of the reactor was observed in the reactor mock-up. The moving velocity and distance of descending particles/sands from the top to bottom of fluidized bed were measured by the visual observation of the colored tracer particles on outside wall of the reactor.

  20. Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Khedr, A.; Abdel-Latif, Salwa H. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt); Abdel-Hadi, Eed A. [Benha Univ., Cairo (Egypt). Shobra Faculty of Engineering; D' Auria, F. [Pisa Univ. (Italy)

    2016-03-15

    In an attempt to understand the built-up of natural circulation in MTR pool type upward flow research reactors after loss of power, an experimental test rig was built to simulate the loop of natural circulation in MTR reactors. The test rig consisting of two vertically oriented branches, in one of them the core is simulated by two rectangular, electrically heated, parallel channels. The other branch simulates the part of the return pipe that participates in the development of core natural circulation. In the first phase of the work, many experimental runs at different conditions of channel's power and branch's initial temperatures are performed. The channel's coolant and surface temperatures were measured. The measurements and their interpretation were published by the first three authors. In the present work the thermal hydraulic behavior of the test rig is complemented by theoretical analysis using RELAP5 Mod 3.3 system code. The analysis consisting of two parts; in the first part RELAP5 model is validated against the measured values and in the second part some of the other not measured hydraulic parameters are predicted and analyzed. The test rig is typically nodalized and an input dick is prepared. In spite of the low pressure of the test rig, the results show that RELAP5 qualitatively predicts the thermal hydraulic behaviour and the accompanied phenomenon of flow inversion of such facilities. Quantitatively, there is a difference between the predicted and measured values especially the channel's surface temperature. This difference may be return to the uncertainties in initial conditions of experimental runs, the position of the thermocouples which buried inside the heat structure, and the heat transfer package in RELAP5.

  1. Topical report: Natural convection shutdown heat removal test facility (NSTF) evaluation for generating additional reactor cavity cooling system (RCCS) data.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C.P.; Lomperski, S.; Aeschlimann, R.W.; Pointer, D.; Nuclear Engineering Division

    2005-09-01

    As part of the Department of Energy (DOE) Generation IV roadmapping activity, the Very High Temperature gas cooled Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept. One of the key passive safety features of the VHTR is the potential for decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-cooled RCCS concept is notably similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that was developed for the General Electric PRISM sodium-cooled fast reactor. As part of the DOE R&D program that supported the development of this fast reactor concept, the Natural Convection Shutdown Heat Removal Test Facility (NSTF) was developed at ANL to provide proof-of-concept data for the RVACS under prototypic natural convection flow, temperature, and heat flux conditions. Due to the similarity between RVACS and the RCCS, current VHTR R&D plans call for the utilization of the NSTF to provide RCCS model development and validation data, in addition to supporting design validation and optimization activities. Both air-cooled and water-cooled RCCS designs are to be included. In support of this effort, ANL has been tasked with the development of an engineering plan for mechanical and instrumentation modifications to NSTF to ensure that sufficiently detailed temperature, heat flux, velocity and turbulence profiles are obtained to adequately qualify the codes under the expected range of air-cooled RCCS flow conditions. Next year, similar work will be carried out for the alternative option of a water-cooled RCCS design. Analysis activities carried out in support of this experiment planning task have shown that: (a) in the RCCS, strong

  2. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Timothy A. Hyde

    2012-06-01

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  3. Development of in-service inspection system for core support graphite structures in the high temperature engineering test reactor (HTTR)

    Energy Technology Data Exchange (ETDEWEB)

    Sumita, Junya; Hanawa, Satoshi; Kikuchi, Takayuki; Ishihara, Masahiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-03-01

    Visual inspection of core support graphite structures using TV camera as in-service inspection and measurement of material characteristics using surveillance test specimens are planned in the High Temperature Engineering Test Reactor (HTTR) to confirm structural integrity of the core support graphite structures. For the visual inspection, in-service inspection system developed from September 1996 to June 1998, and pre-service inspection using the system was carried out. As the result of the pre-service inspection, it was validated that high quality of visual inspection with TV camera can be carried out, and also structural integrity of the core support graphite structures at the initial stage of the HTTR operation was confirmed. (author)

  4. ROSA-III base test series for a large break loss-of-coolant accident in a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tasaka, K.; Abe, N.; Anoda, Y.; Koizumi, Y.; Shiba, M.

    1982-05-01

    The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. It is confirmed from the experimental results obtained so far that the ROSA-III test facility can simulate major aspects of a BWR LOCA, such as boiling transition by lowering of the mixture level in the core, rewetting by the lower plenum flashing, and final quenching by the ECCS. The overall agreement between the calculated results by the RELAP5/ MOD0 code and the experimental results is good; however, the calculated lower plenum flashing rewetted the whole core and the calculated cladding temperature considerably underpredicts the measured value at the upper part of the core.

  5. Radiation-induced electrical degradation experiments in the Japan materials testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Farnum, E.; Scharborough, K. [Los Alamos National Lab., NM (United States); Shikama, Tatsuo [and others

    1995-04-01

    The objective of this experiment is to determine the extent of degradation during neutron irradiation of electrical and optical properties of candidate dielectric materials. The goals are to identify promising dielectrics for ITER and other fusion machines for diagnostic applications and establish the basis for optimization of candidate materials. An experiment to measure radiation-induced electrical degradation (REID) in sapphire and MgO-insulated cables was conducted at the JMTR light water reactor. The materials were irradiated at about 260 {degree}C to a fluence of 3{times}10{sup 24} n/m{sup 2} (E>1 MeV) with an applied DC electric field between 100 kV/m and 500 kV/m.

  6. NE TIGER Premieres New Hua Fu At Bird’s Nest

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    On the evening of September 25, China’s leading luxury fashion brand NE TIGER presented its premiere fashion show of Hua Fu(Chinese national dress) at a concert of superstars from China, Japan and South Korea,

  7. 3D thermal hydraulic simulation of the hot channel of a typical material testing reactor under normal operation conditions

    Energy Technology Data Exchange (ETDEWEB)

    El-Morshedy, Salah El-Din; Salama, Amgad [Atomic Energy Authority, Cairo (Egypt). Reactors Dept.

    2010-09-15

    The hot channel in a typical Material Testing Reactor (MTR) is subjected to 3D simulation. Because of the existence of similarity planes, only a quarter of the hot channel including meat thickness, clad, and coolant channel is considered for CFD analysis using the FLUENT code. For the simulation, steady state normal operation regime at the reactor nominal power is assumed. In order to build confidence in our modeling approach, the results obtained in this work are compared with those obtained from the one-dimensional simulation code, MTRTHA. That is, modified variables were generated in order to match those obtained by MTRTHA and to allow comparisons. Quite good agreement is generally observed, however, the maximum clad surface temperature predicted by the 3D calculations, located at the clad mid-width, is higher than the 1D prediction by about 8 C but still below the onset of subcooled boiling by adequate safety margin. The results show quite interesting 3D patterns in both the flow field and the heat transfer. Temperature profiles, velocity profiles and contours are all presented to highlight the essential 3D features of this system. (orig.)

  8. Fabrication and integrity test preparation of HIP-joined W and ferritic-martensitic steel mockups for fusion reactor development

    Science.gov (United States)

    Lee, Dong Won; Shin, Kyu In; Kim, Suk Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Choi, Bo Guen; Moon, Se Youn; Hong, Bong Guen

    2014-10-01

    Tungsten (W) and ferritic-martensitic steel (FMS) as armor and structural materials, respectively, are the major candidates for plasma-facing components (PFCs) such as the blanket first wall (BFW) and the divertor, in a fusion reactor. In the present study, three W/FMS mockups were successfully fabricated using a hot isostatic pressing (HIP, 900 °C, 100 MPa, 1.5 hrs) with a following post-HIP heat treatment (PHHT, tempering, 750 °C, 70 MPa, 2 hrs), and the W/FMS joining method was developed based on the ITER BFW and the test blanket module (TBM) development project from 2004 to the present. Using a 10-MHz-frequency flat-type probe to ultrasonically test of the joint, we found no defects in the fabricated mockups. For confirmation of the joint integrity, a high heat flux test will be performed up to the thermal lifetime of the mockup under the proper test conditions. These conditions were determined through a preliminary analysis with conventional codes such as ANSYS-CFX for thermal-hydraulic conditions considering the test facility, the Korea heat load test facility with an electron beam (KoHLT-EB), and its water coolant system at the Korea Atomic Energy Research Institute (KAERI).

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

    Science.gov (United States)

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

    2017-08-01

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

  10. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (CAPSULE 2) of Ulchin Nuclear Power Plant Unit 3

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2006-12-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.674E 18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.920 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.913E 18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 9.249E 18 and 1.834E 19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  11. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (Capsule 2) of Ulchin Nuclear Power Plant Unit 4

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2007-04-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin Unit 4 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.306E+18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.918 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.615E+18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 8.478E+18 and 1.673E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin Unit 4 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  12. Acceleration Test Method for Failure Prediction of the End Cap Contact Region of Sodium Cooled Fast Reactor Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Hyun-Seung; Lee, Kang-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-05-15

    This paper reports the results of an acceleration test to predict the contact-induced failure that could occur at the cylinder-to-hole joint for the fuel rod of a sodium-cooled fast reactor (SFR). To incorporate the fuel life of the SFR currently under development at KAERI (around 35,000 h), the acceleration test method of reliability engineering was adopted in this work. A finite element method was used to evaluate the flow-induced vibration frequency and amplitude for the test parameter values. Five specimens were tested. The failure criterion during the life of the SFR fuel was applied. The S-N curve of the HT-9, the material of concern, was used to obtain the acceleration factor. As a result, a test time of 16.5 h was obtained for each specimen. It was concluded that the B{sub 0.004} life would be guaranteed for the SFR fuel rods with 99% confidence if no failure was observed at any of the contact surfaces of the five specimens.

  13. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  14. Modeling and Testing of Non-Nuclear, Highpower Simulated Nuclear Thermal Rocket Reactor Elements

    Science.gov (United States)

    Kirk, Daniel R.

    2005-01-01

    When the President offered his new vision for space exploration in January of 2004, he said, "Our third goal is to return to the moon by 2020, as the launching point for missions beyond," and, "With the experience and knowledge gained on the moon, we will then be ready to take the next steps of space exploration: human missions to Mars and to worlds beyond." A human mission to Mars implies the need to move large payloads as rapidly as possible, in an efficient and cost-effective manner. Furthermore, with the scientific advancements possible with Project Prometheus and its Jupiter Icy Moons Orbiter (JIMO), (these use electric propulsion), there is a renewed interest in deep space exploration propulsion systems. According to many mission analyses, nuclear thermal propulsion (NTP), with its relatively high thrust and high specific impulse, is a serious candidate for such missions. Nuclear rockets utilize fission energy to heat a reactor core to very high temperatures. Hydrogen gas flowing through the core then becomes superheated and exits the engine at very high exhaust velocities. The combination of temperature and low molecular weight results in an engine with specific impulses above 900 seconds. This is almost twice the performance of the LOX/LH2 space shuttle engines, and the impact of this performance would be to reduce the trip time of a manned Mars mission from the 2.5 years, possible with chemical engines, to about 12-14 months.

  15. PITR: a small-aspect-ratio, small-major-radius ignition test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jassby, D.L.; Bolton, R.A.; Brown, D.I.

    1978-05-01

    The principal objectives of the PITR are to demonstrate the attainment of thermonuclear ignition in D-T, and to develop optimal start-up methods for tokamak power reactors. The design approach is based on minimizing dependence on a central transformer core, which thereby results in a machine of small aspect ratio (A approximately 2 to 2.5) and smaller major radius (R/sub 0/ approximately 2.8 m). Current induction is achieved by a combination of ''leaky OH'' coils, equilibrium-field flux swing, a small central solenoid, and compression. Impurity control is effected by a bundle divertor during the beam-heating phase, and by a cold plasma blanket during the burn. The vacuum vessel is constructed of thin-gauge, double-wall titanium alloy. Sixteen normal-copper TF coils of the compound constant-tension type enable low-stress operation at B/sub max/ = 12.5 T.

  16. Development and testing of bioelectrochemical reactors converting wastewater organics into hydrogen peroxide.

    Science.gov (United States)

    Modin, Oskar; Fukushi, Kensuke

    2012-01-01

    In a bioelectrochemical system, the energy content in dissolved organic matter can be used to power the production of hydrogen peroxide (H(2)O(2)), which is a potentially useful chemical at wastewater treatment plants. H(2)O(2) can be produced by the cathodic reduction of oxygen. We investigated four types of gas-diffusion electrodes (GDEs) for this purpose. A GDE made of carbon nanoparticles bound with 30% polytetrafluoroethylene (PTFE) (wt./wt.C) to a carbon fiber paper performed best and catalyzed H(2)O(2) production from oxygen in air with a coulombic efficiency of 95.1%. We coupled the GDE to biological anodes in two bioelectrochemical reactors. When the anodes were fed with synthetic wastewater containing acetate they generated a current of up to ∼0.4 mA/mL total anode compartment volume. H(2)O(2) concentrations of ∼0.2 and ∼0.5% could be produced in 5 mL catholyte in 9 and 21 h, respectively. When the anodes were fed with real wastewater, the generated current was ∼0.1 mA/mL and only 84 mg/L of H(2)O(2) was produced.

  17. Further development of remote testing of submerged bolts and screws in reactors; Weiterentwicklung der ferngesteuerten Schrauben- und Stiftpruefung unter Wasser in Reaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Mohr, F.; Schirner, G.; Meier, R.; Wiesinger, W. [intelligeNDT Systems und Services, Erlangen (Germany)

    2007-07-01

    Since the eighties, intelligeNDT has been carrying out ultrasonic tests of bolts in reactor containments and pressure vessels both in Germany and abroad. The ultrasonic equipment used belonged to the SAPHIR/SAPHIRplus line. The recording and online evaluation software was adapted to the test requirements and optimized for high test rates and quality-assured documentation. As test manipulator, the 'SUSI' submarine by AREVA NP was used with good results. (orig.)

  18. Identifying subassemblies by ultrasound to prevent fuel handling error in sodium fast reactors: First test performed in water

    Energy Technology Data Exchange (ETDEWEB)

    Paumel, Kevin; Lhuillier, Christian [CEA, DEN, Nuclear Technology Department, F-13108 Saint-Paul-lez-Durance, (France)

    2015-07-01

    Identifying subassemblies by ultrasound is a method that is being considered to prevent handling errors in sodium fast reactors. It is based on the reading of a code (aligned notches) engraved on the subassembly head by an emitting/receiving ultrasonic sensor. This reading is carried out in sodium with high temperature transducers. The resulting one-dimensional C-scan can be likened to a binary code expressing the subassembly type and number. The first test performed in water investigated two parameters: width and depth of the notches. The code remained legible for notches as thin as 1.6 mm wide. The impact of the depth seems minor in the range under investigation. (authors)

  19. Inert matrix fuel performance during the first two irradiation cycles in a test reactor: comparison with modelling results

    Science.gov (United States)

    Hellwig, Ch.; Kasemeyer, U.

    2003-06-01

    In the inert matrix fuel (IMF) type investigated at Paul Scherrer Institut, plutonium is dissolved in the yttrium stabilised zirconium oxide (YSZ), a highly radiation resistant cubic phase with additions of erbium as burnable poison for reactivity control. A first irradiation experiment of YSZ based IMF is ongoing in the OECD Material Test Reactor in Halden together with mixed oxide fuel. The results of the first two cycles for IMF to a burnup of some 105 kW d cm -3 are presented and the modelling results in comparison with the experimental results are shown. A first approximation for a simple swelling model for this YSZ based IMF can be given. Possible fission gas release mechanisms are briefly discussed. The implications of the modelling results are discussed.

  20. Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin Allen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, Sebastian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gleicher, Frederick Nathan [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, Mark David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition, this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.

  1. Preliminary Multirod Burst Test Program results and implications of interest to reactor safety evaluations

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, R.H. (comp.)

    1978-01-01

    The Multirod Burst Test (MRBT) Program, in progress at Oak Ridge National Laboratory, is investigating LWR cladding deformation in single- and multirod test arrays under conditions representative of reflood and refill phases of a LOCA. In these tests internally pressurized, unirradiated Zircaloy-4 tubes containing electrically heated fuel simulators are tested to failure in a low-pressure, superheated-steam environment. The tubes are ''uniformly'' heated over a 915-mm length; the simulator pressure, due to the small enclosed gas volume, also varies with temperature (and deformation) during the test. Two 4 x 4 multirod tests (B-1 and B-2), one with and one without the shroud being heated, have been conducted with a bundle heating rate of approx. 29/sup 0/C/sec; initial pressure conditions for these tests were selected to cause failure at about 860/sup 0/C. An additional 4 x 4 array (B-3) was tested using a bundle heating rate of approx. 10/sup 0/C/sec; the shroud was also heated in this test. Initial conditions were adjusted to cause failure at approx. 760/sup 0/C. Posttest examination (including flow tests) of the B-1 and B-2 test arrays, is essentially complete, and pertinent data are included in this summary.

  2. Fabrication of DUPIC fuel pellets for irradiation testing in the NRU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Ki; Kim, S. S.; Park, K. I. [and others

    2003-12-01

    In this project, DUPIC fuel manufacturing processes were established, and the irradiation test of DUPIC fuel at NRU in Canada was planned for the evaluation of DUPIC fuel performance. To establish manufacturing processes satisfying the requirements of NRU irradiation test, pre-qualification test and qualification test were performed. As a result of the qualification test, the DUPIC pellet fabrication processes were qualified and accepted by AECL. 8 batches of experiments were performed to fabricate 375 DUPIC pellets satisfying the requirements of NRU irradiation test under control of the quality assurance manual complying with CAN3-Z299.2-85. Sintered densities of the fabricated DUPIC pellets ranged from 10.26 g/cm{sup 3} to 10.43 g/cm{sup 3}. The DUPIC pellets have been stored in a box filled with helium gas. The pellets will be used to fabricate DUPIC elements for the irradiation testing.

  3. Thick Films acoustic sensors devoted to MTR environment measurements. Thick Films acoustic sensors devoted to Material Testing Reactor environment measurements

    Energy Technology Data Exchange (ETDEWEB)

    Very, F.; Rosenkrantz, E.; Combette, P.; Ferrandis, J.Y. [University Montpellier, IES, UMR 5214, F-34000, Montpellier (France); CNRS, IES, UMR 5214, F-34000, Montpellier (France); Fourmentel, D.; Destouches, C.; Villard, J.F. [CEA, DEN, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul lez Durance (France)

    2015-07-01

    The development of advanced instrumentation for in-pile experiments in Material Testing Reactor constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. An acoustic method for fission gas release detection was tested with success during a first experiment called REMORA 3 in 2010 and 2011, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was lead at OSIRIS reactor (CEA Saclay, France). The maximal temperature on the sensor during the irradiation was about 150 deg. C. In this paper we present a thick film transducer produce by screen printing process. The screen printing of piezoelectric offers a wide range of possible applications for the development of acoustic sensors and piezoelectric structure for measurements in high temperature environment. We firstly produced a Lead Zirconate Titanate (PZT) based paste composed of Pz27 powder from Ferroperm, CF7575 glass, and organic solvent ESL 400. Likewise a Bismuth Titanate based paste synthesized in our laboratory was produced. With these inks we produced thick film up to 130 μm by screen printing process. Material properties characterizations of these thick-film resonators are essential for device design and applications. The piezoelectric coefficients d33 and pyro-electric P(T) coefficient are investigated. The highest P(T) and d33 are respectively 80 μC.m{sup -2}.K{sup -1} and 130 μC.N{sup -1} for the PZT transducer -which validates the fabrication process-. In view of the development of this transducer oriented for high temperature and irradiation environment, we investigated the electrical properties of the transducers for different ranges of frequencies and temperature - from 20 Hz up to 40 MHz between 30 and 400 deg. C. We highlight the evolution of the impedance response and piezoelectric parameters of screen printed piezoelectric structures on alumina. Shortly an irradiation will be realized in

  4. Parametric study of the energy deposition inside the calorimeter measuring the nuclear heating in Material Testing Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Amharrak, H., E-mail: hicham.amharrak@im2np.fr [Aix Marseille Université, CNRS, Université de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Reynard-Carette, C. [Aix Marseille Université, CNRS, Université de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 Saint Paul lez Durance (France); Carette, M.; Brun, J.; De Vita, C. [Aix Marseille Université, CNRS, Université de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Fourmentel, D.; Villard, J-F. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 Saint Paul lez Durance (France)

    2015-11-01

    The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material and two calorimetric cells. Then these measurements are used for other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. This paper will present simulations with MCNP5 Monte-Carlo transport code (using ENDF/B-VI nuclear data library) to evaluate the nuclear heating inside the calorimeter during irradiation campaigns of the CARMEN-1P mock-up inside OSIRIS reactor periphery (MTR based on Saclay, France). The whole complete geometry of the sensor has been considered. The calculation method corresponds to a calculation in two steps. Consequently, we used as an input source in the model, the neutron and photon spectra calculated in various experimental locations tested during the irradiation campaign (H9, H10, H11, D9). After a description of the differential calorimeter sensor, the MCNP5 model used for the calculations of nuclear heating inside the calorimeter elements is introduced by two quantities: KERMA and energy deposition rate per mass unit. The Charged Particle Equilibrium (CPE) inside the calorimeter elements is studied. The contribution of prompt gamma and neutron is determined. A comparison between this total nuclear heating calculation and the experimental results in a graphite sample will be made. Then parametric studies performed on the influence of the various calorimeter components on the nuclear heating are presented and discussed. The studies of the influence of the nature of materials, the sensor jacket, the source type and the comparison of the results obtained for the two calorimetric cells leads to some proposals for the sensor improvement.

  5. Evaluation of the Start-Up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Annular Core Loadings)

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

    2010-03-01

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The Japanese government approved construction of the HTTR in the 1989 fiscal year budget; construction began at the Oarai Research and Development Center in March 1991 and was completed May 1996. Fuel loading began July 1, 1998, from the core periphery. The first criticality was attained with an annular core on November 10, 1998 at 14:18, followed by a series of start-up core physics tests until a fully-loaded core was developed on December 16, 1998. Criticality tests were carried out into January 1999. The first full power operation with an average core outlet temperature of 850ºC was completed on December 7, 2001, and operational licensing of the HTTR was approved on March 6, 2002. The HTTR attained high temperature operation at 950 ºC in April 19, 2004. After a series of safety demonstration tests, it will be used as the heat source in a hydrogen production system by 2015. Hot zero-power critical, rise-to-power, irradiation, and safety demonstration testing , have also been performed with the HTTR, representing additional means for computational validation efforts. Power tests were performed in steps from 0 to 30 MW, with various tests performed at each step to confirm

  6. Research on friction coefficient of nuclear Reactor Vessel Internals Hold Down Spring: Stress coefficient test analysis method

    Energy Technology Data Exchange (ETDEWEB)

    Linjun, Xie, E-mail: linjunx@zjut.edu.cn [College of Mechanical Engineering, Zhejiang University of Technology, Hangzhou 310014 (China); Guohong, Xue; Ming, Zhang [Shanghai Nuclear Engineering Research & Design Institute, Shanghai 200233 (China)

    2016-08-01

    Graphical abstract: HDS stress coefficient test apparatus. - Highlights: • This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. • The mathematical relation between the load and the strain is obtained about the HDS, and the mathematical model of the stress coefficient and the friction coefficient is established. So, a set of test apparatuses for obtaining the stress coefficient is designed according to the model scaling criterion and the friction coefficient of the K1000 HDS is calculated to be 0.336 through the obtained stress coefficient. • The relation curve between the theoretical load and the friction coefficient is obtained through analysis and indicates that the change of the friction coefficient f would influence the pretightening load under the condition of designed stress. The necessary pretightening load in the design process is calculated to be 5469 kN according to the obtained friction coefficient. Therefore, the friction coefficient and the pretightening load under the design conditions can provide accurate pretightening data for the analysis and design of the reactor HDS according to the operations. - Abstract: This paper performs mathematic deduction to the physical model of Hold Down Spring (HDS), establishes a mathematic model of axial load P and stress, stress coefficient and friction coefficient and designs a set of test apparatuses for simulating the pretightening process of the HDS for the first time according to a model similarity criterion. By carrying out tests and researches through a stress testing technique, P–σ curves in loading and unloading processes of the HDS are obtained and the stress coefficient k{sub f} of the HDS is obtained. So, the

  7. Characterization of the relocated and dispersed fuel in the Halden reactor project LOCA tests based on gamma scan data

    Energy Technology Data Exchange (ETDEWEB)

    Brankov, Vladimir, E-mail: vladimir.brankov@psi.ch [Paul Scherrer Institut, 5232 Villigen PSI (Switzerland); École Polytechnique Fédérale de Lausanne, CH-1015 Lausanne (Switzerland); Khvostov, Grigori; Mikityuk, Konstantin [Paul Scherrer Institut, 5232 Villigen PSI (Switzerland); Pautz, Andreas [Paul Scherrer Institut, 5232 Villigen PSI (Switzerland); École Polytechnique Fédérale de Lausanne, CH-1015 Lausanne (Switzerland); Wiesenack, Wolfgang [Institutt For Energiteknikk OECD Halden Reactor Project, P.O. Box 173, Halden 1751 (Norway)

    2016-04-15

    Highlights: • We propose method to estimate dispersed fuel based on gamma scan data. • Analysis to determine the origin of relocated and dispersed fuel in Halden LOCA tests. • Useful data is gathered for code validation. • Suggestions are discussed to improve the quality of gamma scan data at Halden. • Dispersed and relocated material is a mixture of fuel from pellet periphery and bulk. - Abstract: The on-going Loss-of-Coolant Accident (LOCA) test program at the OECD Halden Reactor Project (HRP) conducts in-house gamma scanning as standard post-irradiation examination (PIE) procedure on Light Water Reactor (LWR) fuel rods. One of the primary objectives of the program is to investigate fuel relocation into the balloon region and fuel dispersal through the cladding rupture opening after burst. A simple model called Gamma Transport Model was formulated for the purpose of interpretation of fuel relocation based on the gamma scan data. Fuel relocation may have a strong effect on the linear heat generation rate at the balloon due to, firstly, increase in linear fuel density, and secondly due to differences in burn-up and local heat generation rate at the periphery and bulk of the pellet. For this analysis, a pair of short-lived isotopes with very different fission product yields for {sup 235}U and {sup 239}Pu is selected from the gamma scan spectrum. The intention is to use the difference in the ratio of their concentrations in the balloon region to qualitatively make conclusion on the fuel relocation. As a separate outcome, the same analysis can be applied to the dispersed fuel region and to draw conclusion on its origin (pellet rim or bulk). The Gamma Transport Model is validated against a special (non-destructive) case from the Halden LOCA test program and then applied for the analysis of selected tests. In addition, a methodology is presented for estimation of the amount of dispersed fuel from the LOCA tests based on the gamma scan data. Currently, at

  8. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  9. Strategies for injury prevention in Brazilian football: Perceptions of physiotherapists and practices of premier league teams.

    Science.gov (United States)

    Meurer, Maurício Couto; Silva, Marcelo Faria; Baroni, Bruno Manfredini

    2017-07-25

    To describe the physiotherapists perceptions and the current practices for injury prevention in elite football (soccer) clubs in Brazil. Cross-sectional study. Group of Science in Sports & Exercise, Federal University of Healthy Sciences of Porto Alegre (Brazil). 16 of the 20 football clubs involved in the Brazilian premier league 2015. Physiotherapists answered a structured questionnaire. Most physiotherapists (∼88%) were active in design, testing and application of prevention programs. Previous injury, muscle imbalance, fatigue, hydration, fitness, diet, sleep/rest and age were considered "very important" or "important" injury risk factors by all respondents. The methods most commonly used to detect athletes' injury risk were: monitoring of biochemical markers (100% of teams), isokinetic dynamometry (81%), questionnaires (75%), functional movement screen (56%), fleximetry (56%) and horizontal jump tests (50%). All clubs used strength training, functional training, core exercises and balance/proprioception exercises in their injury prevention program; and Nordic hamstring exercise and other eccentric exercises were used by 94% of clubs. "FIFA 11+" prevention program was adapted by 88% of clubs. Physiotherapists perceptions and current practices of injury prevention within Brazilian elite football clubs were similar to those employed in developed countries. There remains a gap between clinical practice and scientific evidence in high performance football. Copyright © 2017 Elsevier Ltd. All rights reserved.

  10. Long term out-of-pile thermocouple tests in conditions representative for nuclear gas-cooled high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Laurie, M. [European Commission, Joint Research Centre, Inst. for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Fourrez, S. [THERMOCOAX SAS, BP 26, Planquivon, F-61438 Flers Cedex (France); Fuetterer, M. A.; Lapetite, J. M. [European Commission, Joint Research Centre, Inst. for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands)

    2011-07-01

    During irradiation tests at high temperature, failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. To understand and remedy this problem, out-of-pile tests were performed with thermocouples in carburizing atmospheres which can be assumed to be at least locally representative for High Temperature Reactors. The objective was to screen those thermocouples which would consecutively be used under irradiation. Two such screening tests have been performed with a set of thermocouples embedded in graphite (mainly conventional Type N thermocouples and thermocouples with innovative sheaths) in a dedicated furnace with helium flushing. Performance indicators such as thermal drift, insulation and loop resistance were monitored and compared to those from conventional Type N thermocouples. Several parameters were investigated: niobium sleeves, bending, thickness, sheath composition, temperature as well as the chemical environment. After the tests, Scanning Electron Microscopy (SEM) examinations were performed to analyze possible local damage in wires and in the sheath. The present paper describes the two experiments, summarizes results and outlines further work, in particular to further analyze the findings and to select suitable thermocouples for qualification under irradiation. (authors)

  11. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    Energy Technology Data Exchange (ETDEWEB)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  12. Integral and Separate Effects Tests for Thermal Hydraulics Code Validation for Liquid-Salt Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per

    2012-10-30

    The objective of the 3-year project was to collect integral effects test (IET) data to validate the RELAP5-3D code and other thermal hydraulics codes for use in predicting the transient thermal hydraulics response of liquid salt cooled reactor systems, including integral transient response for forced and natural circulation operation. The reference system for the project is a modular, 900-MWth Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a specific type of Fluoride salt-cooled High temperature Reactor (FHR). Two experimental facilities were developed for thermal-hydraulic integral effects tests (IETs) and separate effects tests (SETs). The facilities use simulant fluids for the liquid fluoride salts, with very little distortion to the heat transfer and fluid dynamics behavior. The CIET Test Bay facility was designed, built, and operated. IET data for steady state and transient natural circulation was collected. SET data for convective heat transfer in pebble beds and straight channel geometries was collected. The facility continues to be operational and will be used for future experiments, and for component development. The CIET 2 facility is larger in scope, and its construction and operation has a longer timeline than the duration of this grant. The design for the CIET 2 facility has drawn heavily on the experience and data collected on the CIET Test Bay, and it was completed in parallel with operation of the CIET Test Bay. CIET 2 will demonstrate start-up and shut-down transients and control logic, in addition to LOFC and LOHS transients, and buoyant shut down rod operation during transients. Design of the CIET 2 Facility is complete, and engineering drawings have been submitted to an external vendor for outsourced quality controlled construction. CIET 2 construction and operation continue under another NEUP grant. IET data from both CIET facilities is to be used for validation of system codes used for FHR modeling, such as RELAP5-3D. A set of

  13. The Performance Test for Reactor Coolant Pump (RCP) adopting Variable Restriction Orifice Type Control Valve

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.; Bae, B. U.; Cho, Y. J. and others

    2014-05-15

    The design values of the RCPTF are 17.2 MPa, 343 .deg. C, 11.7 m{sup 3}/s, and 13 MW in the maximum pressure, temperature, flow rate, and electrical power, respectively. In the RCPTF, various types of tests can be performed including a hydraulic performance test to acquire a H-Q curve as well seal transient tests, thrust bearing transient test, cost down test, NPSHR verification test, and so on. After a commissioning startup test was successfully perfomed, mechanical structures are improved including a flow stabilizer and variable restriction orifice. Two- branch pipe (Y-branch) was installed to regulate the flow rate in the range of performance tests. In the main pipe, a flow restrictor (RO: Restriction Orifice) for limiting the maximum flow rate was installed. In the branch pipe line, a globe valve and a butterfly valves for regulating the flow rate was located on the each branch line. When the pressure loss of the valve side is smaller than that of the RO side, the flow rate of valve side was increasing and the flow disturbance was occurred in the lower pipe line. Due to flow disturbnace, it is to cause an error when measuring RCP head and flow measurement of the venturi flow meter installed in the lower main pipe line, and thus leading to a decrease in measurement accuracy as a result. To increase the efficiency of the flow control availability of the test facility, the variable restriction orifice (VRO) type flow control valve was designed and manufactured. In the RCPTF in KAERI, the performance tests and various kinds of transient tests of the RCP were successfully performed. In this study, H-Q curve of the pump using the VRO revealed a similar trend to the result from two ROs. The VRO was confirmed to effectively cover the full test range of the flow rate.

  14. Research on the improvement of nuclear safety -Thermal hydraulic tests for reactor safety system-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Moon Kee; Park, Choon Kyung; Yang, Sun Kyoo; Chun, Se Yung; Song, Chul Hwa; Jun, Hyung Kil; Jung, Heung Joon; Won, Soon Yun; Cho, Yung Roh; Min, Kyung Hoh; Jung, Jang Hwan; Jang, Suk Kyoo; Kim, Bok Deuk; Kim, Wooi Kyung; Huh, Jin; Kim, Sook Kwan; Moon, Sang Kee; Lee, Sang Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-06-01

    The present research aims at the development of the thermal hydraulic verification test technology for the safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. In this research, test facilities simulating the primary coolant system and safety system are being constructed for the design verification tests of the existing and advanced nuclear power plant. 97 figs, 14 tabs, 65 refs. (Author).

  15. 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.

    Energy Technology Data Exchange (ETDEWEB)

    Garner, P. L.; Hanan, N. A. (Nuclear Engineering Division)

    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.

  16. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Jorge [Univ. of Utah, Salt Lake City, UT (United States)

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  17. Model test of reactor vessel wall. Pt. 2. Test performance, measurement and partial evaluation; Modellkoerperversuch zur Reaktortankwand. T. 2; Versuchsdurchfuehrung, Messung und Teilauswertung

    Energy Technology Data Exchange (ETDEWEB)

    Maile, K.; Eckert, W.; Theofel, H.; Purper, H.

    1992-07-01

    Due to test interruption because of cut promotion means, the original objective of the project - verification of reactor wall design - could not be achieved because by that point in time the slabs had not yet failed (DIN 1.4948 = X 6 CrNi 18 11). Considering, however, the elongation curve, in particular that of the faulty slab, failure at an earlier stress cycle value than calculated is highly probable. (orig./HP) [Deutsch] Aufgrund des foerderungsbedingten Abbruchs der Versuche konnte die urspruengliche Zielsetzung des Vorhabens - Verifizierung der Auslegung der Reaktorwand - nicht erreicht werden, da ein Versagen der Platten bis zu diesem Zeitpunkt noch nicht vorlag (DIN 1.4948 = X 6 CrNi 18 11). Betrachtet man jedoch den Dehnungsverlauf insbesondere in der Platte mit Fehlern ist zu vermuten, dass ein Versagen zu einer frueheren Lastspielzahl als berechnet sehr wahrscheinlich ist. (orig./HP)

  18. Predicting Football Matches Results using Bayesian Networks for English Premier League (EPL)

    Science.gov (United States)

    Razali, Nazim; Mustapha, Aida; Yatim, Faiz Ahmad; Aziz, Ruhaya Ab

    2017-08-01

    The issues of modeling asscoiation football prediction model has become increasingly popular in the last few years and many different approaches of prediction models have been proposed with the point of evaluating the attributes that lead a football team to lose, draw or win the match. There are three types of approaches has been considered for predicting football matches results which include statistical approaches, machine learning approaches and Bayesian approaches. Lately, many studies regarding football prediction models has been produced using Bayesian approaches. This paper proposes a Bayesian Networks (BNs) to predict the results of football matches in term of home win (H), away win (A) and draw (D). The English Premier League (EPL) for three seasons of 2010-2011, 2011-2012 and 2012-2013 has been selected and reviewed. K-fold cross validation has been used for testing the accuracy of prediction model. The required information about the football data is sourced from a legitimate site at http://www.football-data.co.uk. BNs achieved predictive accuracy of 75.09% in average across three seasons. It is hoped that the results could be used as the benchmark output for future research in predicting football matches results.

  19. Evaluation of sports nutrition knowledge of New Zealand premier club rugby coaches.

    Science.gov (United States)

    Zinn, Caryn; Schofield, Grant; Wall, Clare

    2006-04-01

    Little is known about if and how team coaches disseminate nutrition information to athletes. In a census survey, New Zealand premier rugby coaches (n = 168) completed a psychometrically validated questionnaire, received by either Internet or standard mail (response rate, 46%), identifying their nutrition advice dissemination practices to players, their level of nutrition knowledge, and the factors determining this level of knowledge. The majority of coaches provided advice to their players (83.8%). Coaches responded correctly to 55.6% of all knowledge questions. An independent t-test showed coaches who imparted nutrition advice obtained a significantly greater score, 56.8%, than those not imparting advice, 48.4% (P = 0.008). One-way ANOVA showed significant relationships between total knowledge score of all coaches and qualifications [F(1,166) = 5.28, P = 0.001], own knowledge rating [F(3,164) = 6.88, P = 0.001] and nutrition training [F(1,166) = 9.83, P = 0.002]. We conclude that these rugby coaches were inadequately prepared to impart nutrition advice to athletes and could benefit from further nutrition training.

  20. Technical Performance Analysis of Iran Premier League Soccer Players in 2012-2013 Season

    Directory of Open Access Journals (Sweden)

    Mohsen Javani

    2015-10-01

    Full Text Available Background and purpose of study : analysis of IRAN premier league soccer players’ technical performance in season 2012-2013, using a computerized match analysis system (Borhan Mobin Development Management Co, IRAN. Material and methods: in this study, data were obtained from 120 players, who performed in competitions 90 minutes. The players were classified into 3 positional roles: defenders, midfielders and forwards. Technical performance variables analysis included: total passes, total successful passes, pass accuracy, total shots; total shots to target, shot accuracy, ball interception and ball losses. The data were statistically analyzed by one-way ANOVA, Kruskal-Wallis, Mann-Whitney U and Tukey post hoc test. Results : The findings of this study showed that players performed about 45 passes per competition. Midfielders and defenders had significantly higher number of passes than forwards. Pass accuracy was about 67% and there were no significant differences between positional roles. Also, the players performed about 0.8 shots per competition, forwards and midfielders had significantly higher number of shots than defenders. Shot accuracy was about 31%; midfielders and forwards had significantly higher shot accuracy than defenders. Forwards showed significantly lower ball interception and higher ball losses than other positions. Conclusion : The result of this study showed that there were significant differences between some technical actions in positional roles. Therefore, coaches can use this information for individualization of training according to playing positions and for optimization of training in the amateur game.

  1. The basic approach to the performance test of the ALIP for GENIV sodium fast reactor

    Science.gov (United States)

    Kim, H. R.; Lee, Y. B.

    2016-08-01

    The annular linear induction electromagnetic pump (ALIP) with the flowrate of 900 L/min and the developed pressure of 4 bar has been designed by electric equivalent circuit analysis. It was fabricated by the consideration of materials compatible to the sodium environment of high temperature. Basic characteristic test of the ALIP was carried out in advance for its installation to the integral effect TEst Loop for safety simuLation and Assessment (STELLA) loop to confirm the sodium-thermo-hydraulic components. The test showed that the magnetic field had been linearly increased when the input current was increased, where input current and voltage had represented linear relation each other. The generated electromagnetic force was proportionate to the square of the applied current. The velocity of the aluminium pipe was proportionally increased when the input current was increased. It was verified that the basic characteristic of the ALIP showed a good accordance with the theoretical calculation.

  2. FY15 Status Report: CIRFT Testing of Spent Nuclear Fuel Rods from Boiler Water Reactor Limerick

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-06-01

    The objective of this project is to perform a systematic study of used nuclear fuel (UNF, also known as spent nuclear fuel [SNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. The additional CIRFT was conducted on three HBR rods (R3, R4, and R5) in which two specimens failed and one specimen was tested to over 2.23 10⁷ cycles without failing. The data analysis on all the HBR UNF rods demonstrated that it is necessary to characterize the fatigue life of the UNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum of tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, ten SNF rod segments from BWR Limerick were tested using ORNL CIRFT, with one under static and nine dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at maximum curvature 4.0 m⁻¹. The specimen did not show any sign of failure in three repeated loading cycles to almost same maximum curvature. Ten cyclic tests were conducted with amplitude varying from 15.2 to 7.1 N·m. Failure was observed in nine of the tested rod specimens. The cycles to failure were

  3. Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D., E-mail: thiagodbtr@gmail.com [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil); Silva, Mário A. B. da, E-mail: mabs500@gmail.com [Departamento de Energia Nuclear (CTG/UFPE), Av. Professor Luiz Freire, 1000, Recife 50740-540, PE (Brazil); Lapa, Celso M.F., E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil)

    2016-01-15

    The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

  4. Development and experimental validation of a calculation scheme for nuclear heating evaluation in the core of the OSIRIS material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Malouch, F. [Saclay Center CEA, DEN/DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette Cedex (France)

    2011-07-01

    The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron-photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation. (authors)

  5. Performance of AGR-1 High-Temperature Reactor Fuel During Post-Irradiation Heating Tests

    Energy Technology Data Exchange (ETDEWEB)

    Morris, Robert Noel [ORNL; Baldwin, Charles A [ORNL; Hunn, John D [ORNL; Demkowicz, Paul [Idaho National Laboratory (INL); Reber, Edward [Idaho National Laboratory (INL)

    2014-01-01

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide TRISO fuel compacts from the AGR-1 experiment has been evaluated at temperatures of 1600 1800 C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4 to 19.1% FIMA have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 10-6 after 300 h at 1600 C or 100 h at 1800 C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 C, and 85Kr release was very low during the tests (particles with breached SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 C in one compact. Post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.

  6. SNTP program reactor design

    Science.gov (United States)

    Walton, Lewis A.; Sapyta, Joseph J.

    1993-06-01

    The Space Nuclear Thermal Propulsion (SNTP) program is evaluating the feasibility of a particle bed reactor for a high-performance nuclear thermal rocket engine. Reactors operating between 500 MW and 2,000 MW will produce engine thrusts ranging from 20,000 pounds to 80,000 pounds. The optimum reactor arrangement depends on the power level desired and the intended application. The key components of the reactor have been developed and are being tested. Flow-to-power matching considerations dominate the thermal-hydraulic design of the reactor. Optimal propellant management during decay heat cooling requires a three-pronged approach. Adequate computational methods exist to perform the neutronics analysis of the reactor core. These methods have been benchmarked to critical experiment data.

  7. As-Run Physics Analysis for the UCSB-1 Experiment in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, Joseph Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The University of California Santa Barbara (UCSB) -1 experiment was irradiated in the A-10 position of the ATR. The experiment was irradiated during cycles 145A, 145B, 146A, and 146B. Capsule 6A was removed from the test train following Cycle 145A and replaced with Capsule 6B. This report documents the as-run physics analysis in support of Post-Irradiation Examination (PIE) of the test. This report documents the as-run fluence and displacements per atom (DPA) for each capsule of the experiment based on as-run operating history of the ATR. Average as-run heating rates for each capsule are also presented in this report to support the thermal analysis.

  8. RELAP5-3D version 4.0.3: installation and tests for applications to space reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lobo, Paulo D.C.; Braz Filho, Francisco A.; Borges, Eduardo M.; Guimaraes, Lamartine N.F., E-mail: plobo.a@uol.com.br, E-mail: fbraz@ieav.cta.br, E-mail: eduardo@ieav.cta.br, E-mail: guimarae@ieav.cta.br [Instituto de Estudos Avancados (IEAv), Sao Jose dos Campos, SP (Brazil); Sabundjian, Gaiane, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    To attend the TERRA project (Tecnologia de Reatores Rapidos Avancados), currently conducted by the Nuclear Energy Division (ENU) of the IEAv, this work presents the RELAP5-3D, Version 4.0.3, prepared in July 12, 2012, also known as r3d403is, received recently by the IEAv from the Idaho National Laboratory (INL). This version of RELAP5-3D is configured for the International User Group source Code Group and is developed and maintained at the INL for the US Department of Energy. RELAP5-3D, the latest in the series of RELAP5 codes, is a highly generic code that, in addition to calculating the behavior of a reactor coolant system during a transient, can be used for simulation of a wide variety of hydraulic and thermal transients in both nuclear and nonnuclear systems involving mixtures of vapor, liquid, noncondensable gases, and nonvolatile solute. Enhancements include all features and models previously available in the ATHENA configuration version of the code which are as follows: addition of new work fluids and a magneto-hydrodynamic mode. Following the instructions from the README file, the RELAP5-3D, version 4.0.3 was installed creating the necessaries subdirectories, by using the LINUX platform and applying both Intel Fortran 95 and C-language compilers. Many input examples were executed and the same results were observed as compared to the received documentation. A sample of the Edwards-O'Brien test was evaluated to verify if the code could simulate a LOCA type accident properly. The test executed by the RELAP5-3D demonstrated good agreement with test data including a new output involving the mass flow during the test. (author)

  9. Pre-qualification experiments of DUPIC fuel pellets for irradiation testing in the NRU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Ki; Kim, S. S.; Lee, J. W. and others

    2002-02-01

    DUPIC fuel manufacturing technologies and processes have been developed at DFDF(DUPIC Fuel Development Facility, IMEF M6). Using DUPIC powder prepared by the oxidation and reduction processes, the DUPIC fuel pellets and mini-elements were fabricated for the irradiation test and performance evaluation at HANARO. In this study, the irradiation test was planned for the performance evaluation of DUPIC fuel pellets and elements at NRU. To establish fabrication process satisfying the requirements of NRU irradiation test, sintered DUPIC pellets were fabricated with a variety of process parameters involving compaction pressure and characterized by the inspection system. As a result of the experiment, DUPIC pellets with 12.19 mm of diameter, 10.37{approx}10.45 g/cm{sup 3} of sintered density, and less than Ra 0.8{mu}m of surface roughness have been successfully fabricated at hot cell. The optimum DUPIC pellet fabrication process satisfying the requirements of NRU irradiation has been established based on the result of this experiment.

  10. Development of Mini-Compact Tension Test Method for Determining Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-05-01

    Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Validation of the mini compact tension specimen (mini-CT) geometry has been performed on previously well characterized Midland beltline Linde 80 (WF-70) weld in the unirradiated condition. It was shown that the fracture toughness transition temperature, To, measured by these Mini-CT specimens is almost the same as To value that was derived from various larger fracture toughness specimens. Moreover, an International collaborative program has been established to extend the assessment and validation efforts to irradiated Linde 80 weld metal. The program is underway and involves the Oak Ridge National Laboratory (ORNL), Central Research Institute for Electrical Power Industry (CRIEPI), and Electric Power Research Institute (EPRI). The irradiated Mini-CT specimens from broken halves of previously tested Charpy

  11. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

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

  12. Space Nuclear Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-06

    We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

  13. New Sensors for the Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Joy L. Rempe; Darrell L. Knudson; Keith G. Condie; Joshua E. Daw; Heng Ban; Brandon Fox; Gordon Kohse

    2009-06-01

    A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the selection strategy of what instrumentation is needed, and the program generated for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available to users of the ATR NSUF with data from irradiation tests using these sensors. In addition, progress is reported on current research efforts to provide users advanced methods for detecting temperature, fuel thermal conductivity, and changes in sample geometry.

  14. Wind Tunnel Testing Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — NASA Ames Research Center is pleased to offer the services of our premier wind tunnel facilities that have a broad range of proven testing capabilities to customers...

  15. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Younggwang nuclear power plant unit1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 1 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.555E+18, 1.662E+19, 3.358E+19, and 4.521E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.859 for the 1st through 4th testing and the calculational uncertainty, 11.80% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.551E+19 n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.929E+19, 4.880E+19, 5.831E+19 and 6.782E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 1 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 4 refs., 41 figs., 35 tabs. (Author)

  16. Final report for the 5th surveillance test of the reactor pressure vessel material (capsule Y) of Yonggwang Nuclear Power Plant unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sam Lai; Kim, ByoungChul; Chang, Kee Ok (and others)

    2006-02-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X, W and Y are 5.777E+18, 1.5371E+19, 3.7634E+19, 4.3045E+19, and 4.8662E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.953 for the 1st through 5th testing and the calculational uncertainty,7.2% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.659E+19n/cm{sup 2} based on the end of 13th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 48, 56 and 64EFPY would reach 3.625E+19, 5.293E+19, 6.127E+19 and 6.960E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the Pressurized Thermal Shock(PTS) during the operation until design life.

  17. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Yonggwang nuclear power plant unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-02-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.762E+18, 1.5391E+19, 3.5119E+19, and 4.2610E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.899 for the 1st through 4th testing and the calculational uncertainty, 12.3% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.357E+19 n/cm{sup 2} based on the end of 11th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.525E+19, 4.337E+19, 5.148E+19 and 5.960E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 42 tabs. (Author)

  18. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  19. Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

    2015-07-01

    The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)

  20. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  1. Microinstability properties of negative magnetic shear discharges in the Tokamak Fusion Test Reactor and DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Rewoldt, G.; Tang, W.M. [Princeton Univ., NJ (United States). Princeton Plasma Physics Lab.; Lao, L.L. [General Atomics, San Diego, CA (United States)

    1997-03-01

    The microinstability properties of discharges with negative (reversed) magnetic shear in the Tokamak Fusion Test Reactor (TFTR) and DIII-D experiments with and without confinement transitions are investigated. A comprehensive kinetic linear eigenmode calculation employing the ballooning representation is employed with experimentally measured profile data, and using the corresponding numerically computed magnetohydrodynamic (MHD) equilibria. The instability considered is the toroidal drift mode (trapped-electron-{eta}{sub i} mode). A variety of physical effects associated with differing q-profiles are explained. In addition, different negative magnetic shear discharges at different times in the discharge for TFTR and DIII-D are analyzed. The effects of sheared toroidal rotation, using data from direct spectroscopic measurements for carbon, are analyzed using comparisons with results from a two-dimensional calculation. Comparisons are also made for nonlinear stabilization associated with shear in E{sub r}/RB{sub {theta}}. The relative importance of changes in different profiles (density, temperature, q, rotation, etc.) on the linear growth rates is considered.

  2. Process Knowledge Summary Report for Advanced Test Reactor Complex Contact-Handled Transuranic Waste Drum TRA010029

    Energy Technology Data Exchange (ETDEWEB)

    B. R. Adams; R. P. Grant; P. R. Smith; J. L. Weisgerber

    2013-09-01

    This Process Knowledge Summary Report summarizes information collected to satisfy the transportation and waste acceptance requirements for the transfer of one drum containing contact-handled transuranic (TRU) actinide standards generated by the Idaho National Laboratory at the Advanced Test Reactor (ATR) Complex to the Advanced Mixed Waste Treatment Project (AMWTP) for storage and subsequent shipment to the Waste Isolation Pilot Plant for final disposal. The drum (i.e., Integrated Waste Tracking System Bar Code Number TRA010029) is currently stored at the Materials and Fuels Complex. The information collected includes documentation that addresses the requirements for AMWTP and applicable sections of their Resource Conservation and Recovery Act permits for receipt and disposal of this TRU waste generated from ATR. This Process Knowledge Summary Report includes information regarding, but not limited to, the generation process, the physical form, radiological characteristics, and chemical contaminants of the TRU waste, prohibited items, and packaging configuration. This report, along with the referenced supporting documents, will create a defensible and auditable record for this TRU waste originating from ATR.

  3. Analytical estimation of control rod shadowing effect for excess reactivity measurement of High Temperature Engineering Test Reactor (HTTR)

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, Masaaki; Yamashita, Kiyonobu; Fujimoto, Nozomu; Nojiri, Naoki; Takeuchi, Mitsuo; Fujisaki, Shingo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tokuhara, Kazumi; Nakata, Tetsuo

    1998-05-01

    The control rod shadowing effect has been estimated analytically in application of the fuel addition method to excess reactivity measurement of High Temperature Engineering Test Reactor (HTTR). The movements of control rods in the procedure of the fuel addition method have been simulated in the analysis. The calculated excess reactivity obtained by the simulation depends on the combinations of measuring control rods and compensating control rods and varies from -10% to +50% in comparison with the excess reactivity calculated from the effective multiplication factor of the core where all control rods are fully withdrawn. The control rod shadowing effect is reduced by the use of plural number of measuring and compensation control rods because of the reduction in neutron flux deformation in the measuring procedure. As a result, following combinations of control rods are recommended; 1) Thirteen control rods of the center, first, and second rings will be used for the reactivity measurement. The reactivity of each control rod is measured by the use of the other twelve control rods for reactivity compensation. 2) Six control rods of the first ring will be used for the reactivity measurement. The reactivity of each control rod is measured by the use of the other five control rods for reactivity compensation. (author)

  4. 2016 Annual Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Ponds

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Michael George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-02-01

    This report describes conditions and information, as required by the state of Idaho, Department of Environmental Quality Reuse Permit I-161-02, for the Advanced Test Reactor Complex Cold Waste Ponds located at Idaho National Laboratory from November 1, 2015–October 31, 2016. The effective date of Reuse Permit I-161-02 is November 20, 2014 with an expiration date of November 19, 2019. This report contains the following information: • Facility and system description • Permit required effluent monitoring data and loading rates • Permit required groundwater monitoring data • Status of compliance activities • Issues • Discussion of the facility’s environmental impacts. During the 2016 permit year, 180.99 million gallons of wastewater were discharged to the Cold Waste Ponds. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest in well USGS-065, which is the closest downgradient well to the Cold Waste Ponds. Sulfate and total dissolved solids concentrations decrease rapidly as the distance downgradient from the Cold Waste Ponds increases. Although concentrations of sulfate and total dissolved solids are significantly higher in well USGS-065 than in the other monitoring wells, both parameters remained below the Ground Water Quality Rule Secondary Constituent Standards in well USGS-065. The facility was in compliance with the Reuse Permit during the 2016 permit year.

  5. Preliminary structural design and thermo-mechanical analysis of helium cooled solid breeder blanket for Chinese Fusion Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Li, Min; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Lv, Zhongliang; Ye, Minyou

    2015-02-15

    Highlights: • A helium cooled solid breeder blanket module was designed for CFETR. • Multilayer U-shaped pebble beds were adopted in the blanket module. • Thermal and thermo-mechanical analyses were carried out under normal operating conditions. • The analysis results were found to be acceptable. - Abstract: With the aim to bridge the R&D gap between ITER and fusion power plant, the Chinese Fusion Engineering Test Reactor (CFETR) was proposed to be built in China. The mission of CFETR is to address the essential R&D issues for achieving practical fusion energy. Its blanket is required to be tritium self-sufficient. In this paper, a helium cooled solid breeder blanket adopting multilayer U-shaped pebble beds was designed and analyzed. Thermo-mechanical analysis of the first wall and side wall combined with breeder unit was carried out for normal operating steady state conditions. The results showed that the maximum temperatures of the structural material, neutron multiplier and tritium breeder pebble beds are 523 °C, 558 °C and 787 °C, respectively, which are below the corresponding limits of 550 °C, 650 °C and 920 °C. The maximum equivalent stress of the structure is under the allowable value with a margin about 14.5%.

  6. Microbial fouling community analysis of the cooling water system of a nuclear test reactor with emphasis on sulphate reducing bacteria.

    Science.gov (United States)

    Balamurugan, P; Joshi, M Hiren; Rao, T S

    2011-10-01

    Culture and molecular-based techniques were used to characterize bacterial diversity in the cooling water system of a fast breeder test reactor (FBTR). Techniques were selected for special emphasis on sulphate-reducing bacteria (SRB). Water samples from different locations of the FBTR cooling water system, in addition to biofilm scrapings from carbon steel coupons and a control SRB sample were characterized. Whole genome extraction of the water samples and SRB diversity by group specific primers were analysed using nested PCR and denaturing gradient gel electrophoresis (DGGE). The results of the bacterial assay in the cooling water showed that the total culturable bacteria (TCB) ranged from 10(3) to 10(5) cfu ml(-1); iron-reducing bacteria, 10(3) to 10(5) cfu ml(-1); iron oxidizing bacteria, 10(2) to 10(3) cfu ml(-1) and SRB, 2-29 cfu ml(-1). However, the counts of the various bacterial types in the biofilm sample were 2-3 orders of magnitude higher. SRB diversity by the nested PCR-DGGE approach showed the presence of groups 1, 5 and 6 in the FBTR cooling water system; however, groups 2, 3 and 4 were not detected. The study demonstrated that the PCR protocol influenced the results of the diversity analysis. The paper further discusses the microbiota of the cooling water system and its relevance in biofouling.

  7. Local transport barrier formation and relaxation in reverse-shear plasmas on the Tokamak Fusion Test Reactor

    Science.gov (United States)

    Synakowski, E. J.; Batha, S. H.; Beer, M. A.; Bell, M. G.; Bell, R. E.; Budny, R. V.; Bush, C. E.; Efthimion, P. C.; Hahm, T. S.; Hammett, G. W.; LeBlanc, B.; Levinton, F.; Mazzucato, E.; Park, H.; Ramsey, A. T.; Schmidt, G.; Rewoldt, G.; Scott, S. D.; Taylor, G.; Zarnstorff, M. C.

    1997-05-01

    The roles of turbulence stabilization by sheared E×B flow and Shafranov shift gradients are examined for Tokamak Fusion Test Reactor [D. J. Grove and D. M. Meade, Nucl. Fusion 25, 1167 (1985)] enhanced reverse-shear (ERS) plasmas. Both effects in combination provide the basis of a positive-feedback model that predicts reinforced turbulence suppression with increasing pressure gradient. Local fluctuation behavior at the onset of ERS confinement is consistent with this framework. The power required for transitions into the ERS regime are lower when high power neutral beams are applied earlier in the current profile evolution, consistent with the suggestion that both effects play a role. Separation of the roles of E×B and Shafranov shift effects was performed by varying the E×B shear through changes in the toroidal velocity with nearly steady-state pressure profiles. Transport and fluctuation levels increase only when E×B shearing rates are driven below a critical value that is comparable to the fastest linear growth rates of the dominant instabilities. While a turbulence suppression criterion that involves the ratio of shearing to linear growth rates is in accord with many of these results, the existence of hidden dependencies of the criterion is suggested in experiments where the toroidal field was varied. The forward transition into the ERS regime has also been examined in strongly rotating plasmas. The power threshold is higher with unidirectional injection than with balance injection.

  8. Virtual accelerator concept, implementation and preliminary test; Accelerateur virtuel Concept, implementation et premier test

    Energy Technology Data Exchange (ETDEWEB)

    Uriot, D.; Duperrier, R

    2006-05-15

    A virtual accelerator is the coupling of a simulation code with the control system of a real machine. 3 operating modes are considered. First, the monitoring mode in which any action on the control system has an impact on both real and virtual machines. This mode allows a direct comparison between simulation results and the real behaviour of the accelerator. Secondly, the flight simulation mode, this mode allows the accelerator operators to simulate the effect of any change in the parameters of the control system before transferring them to the real machine. The main advantage of this mode is to allow the assessment of operating procedures before implementing them on the real machine. The third mode is the automatic steering mode in which the simulation code assumes the reins of the control system of the real machine. This mode allows the making of complex and time-consuming adjustment procedures in an automatic way. TraceWin is a simulation code dedicated to the behaviour of charged-particle beams in a linear accelerator. TraceWin is consistent with the EPICS technology on which the control system of most accelerators is based. A virtual accelerator composed of the SILHI injector combined to the TraceWin code via the EPICS environment has showed its efficiency in the automatic steering mode. (A.C.)

  9. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  10. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  11. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  12. Osiris, an irradiation reactor for material and nuclear fuel testing; Osiris, reacteur d'irradiation pour materiaux et combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Loubiere, S.; Durande-Ayme, P. [CEA Saclay, Div. Nucleaire Energie, Dept. Reacteurs et Nucleaire Service, 91 - Gif-sur-Yvette (France)

    2005-07-01

    Since 1966 the Osiris reactor located at Saclay has been participating in French and international irradiation programs for research and development in the field of nuclear fuel and materials. Today the French atomic commission (Cea) pursues irradiation programs in support of existing reactors, mainly PWR, strengthening its own knowledge and the one of its clients on fuel and material behaviour under irradiation, pertaining to plant life-time issues and high burn-up. For instance important programs have been performed on pressure vessel steel aging, pellet-clad interaction, internal component aging and mox fuel qualification. With the arising of the Generation 4 research and development programs, the Osiris reactor has developed capacities to undertake material and fuel irradiation under high temperature conditions. Routine irradiations such as the doping of silicon or the production of radio-nuclides for medical or imaging purposes are made on a daily basis. The specificities of the Osiris reactor are presented in the first part of this paper while the second part focuses on the experimental devices available in Osiris to perform irradiation in light water reactor conditions and in high temperature reactor conditions and on their associated programs.

  13. Experiment Safety Assurance Package for Mixed Oxide Fuel Irradiation in an Average Power Position (I-24) in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    J. M . Ryskamp; R. C. Howard; R. C. Pedersen; S. T. Khericha

    1998-10-01

    The Fissile Material Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation Test Project Plan details a series of test irradiations designed to investigate the use of weapons-grade plutonium in MOX fuel for light water reactors (LWR) (Cowell 1996a, Cowell 1997a, Thoms 1997a). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons-derived test fuel contains small amounts of gallium (about 2 parts per million). A concern exists that the gallium may migrate out of the fuel and into the clad, inducing embrittlement. For preliminary out-of-pile experiments, Wilson (1997) states that intermetallic compound formation is the principal interaction mechanism between zircaloy cladding and gallium. This interaction is very limited by the low mass of gallium, so problems are not expected with the zircaloy cladding, but an in-pile experiment is needed to confirm the out-of-pile experiments. Ryskamp (1998) provides an overview of this experiment and its documentation. The purpose of this Experiment Safety Assurance Package (ESAP) is to demonstrate the safe irradiation and handling of the mixed uranium and plutonium oxide (MOX) Fuel Average Power Test (APT) experiment as required by Advanced Test Reactor (ATR) Technical Safety Requirement (TSR) 3.9.1 (LMITCO 1998). This ESAP addresses the specific operation of the MOX Fuel APT experiment with respect to the operating envelope for irradiation established by the Upgraded Final Safety Analysis Report (UFSAR) Lockheed Martin Idaho Technologies Company (LMITCO 1997a). Experiment handling activities are discussed herein.

  14. The test channel groups in the Wolsong-3 reactor for the evaluation of PHWR fuel integrity in the two phase flow condition

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Cho, Moon Sung

    2004-07-01

    This report intends to select the test channel groups in the Wolsong-3 reactor for the evaluation of PHWR fuel integrity in the two phase flow condition. First, a detailed work schedule for fuel inspection, fuel loading, storage of the discharged fuel, and in-bay visual and dimensional examination was established on the CIGAR work plan of the Wolsong-3 reactor. According to the schedule, it is necessary to suggest the appropriate test channel groups with two phase flow before the new fuel loading. For the selection of the test channel groups, the distributions of the channel flow rate and the channel exit quality are calculated by the NUCIRC code with the monthly measured operation data of the Wolsong-3 reactor during May 2001 to December 2002. The monthly calculated data could sort out the channels with the channel exit quality greater than 0.2%, in the high power channels with the power higher than 6300 kW and the flow rate greater than 25.5 kg/s, and in the low power channels with the power higher than 4800 kW and less than 5300 kW. Based on the frequency rank of these sorted channels, O16 channel with thirty other channels in the high power region and D17 channel with ten other channels in the low power region are suggested as the test channel groups.

  15. The 2014 presidential elections and their impact on the premier-presidential regime in Romania

    Directory of Open Access Journals (Sweden)

    Ionela Gavril

    2015-03-01

    Full Text Available First, we will demonstrate that, from an institutional perspective, Romania can labeled of premier-presidentialism regime, but the 2004 and 2009 elections have had a strong impact on the type of regime, meaning that several extra-constitutional factors led to the malfunction of the regime. Out of a total of 15 prime-minister nominations made after 1989, 8 can be considered deviations from the premier-presidential regime, their number being larger between 2004-2014 rather than in 1990-2000. The empirical analysis of the 2004-2014 period, highlighted three extra-constitutional factors that that made the premier-presidential regime be, in fact, a malfunctioning one: leadership style, crisis situations and the recent legitimacy of the president versus the parliament. By identifying the factors that influenced the regime type, we can determine some theoretical expectations following the 2014 elections. The success of a premier-presidentialism regime in Romania will be determined by the number of deviations from such a regime registered after the 2014 elections.

  16. Effectiveness of in-season manager changes in English Premier League Football

    NARCIS (Netherlands)

    Besters, Lucas; van Ours, Jan; van Tuijl, Martin

    2016-01-01

    We analyze the performance effects of in-season manager changes in English Premier League football during the seasons 2000/2001–2014/2015. We find that some managerial changes are successful, while others are counterproductive. On average, performance does not improve following a managerial replacem

  17. Vice Premier Zeng Peiyan Meets with Michael Cohrs, CEO of Corporate & Investment Banking of Deutsche Bank

    Institute of Scientific and Technical Information of China (English)

    2004-01-01

    <正>Invited by the CPAFFC, Michael Cohrs, Chief Executive Officer of the Corporate and Investment Banking of Deutsche Bank, visited Beijing from February 26 to 28, 2004. Vice Premier Zeng Peiyan of the State Council met with Cohrs and his party. Deutsche Bank AG, a global multipurpose bank

  18. Effectiveness of in-season manager changes in English Premier League Football

    NARCIS (Netherlands)

    Besters, Lucas; van Ours, Jan; van Tuijl, Martin

    We analyze the performance effects of in-season manager changes in English Premier League football during the seasons 2000/2001–2014/2015. We find that some managerial changes are successful, while others are counterproductive. On average, performance does not improve following a managerial

  19. Profil de l'etudiant du premier cycle des etudes medicales de Lome ...

    African Journals Online (AJOL)

    Profil de l'etudiant du premier cycle des etudes medicales de Lome et sa perception de l'enseignement de l'anatomie. ... Journal de la Recherche Scientifique de l'Universite de Lome ... aux différentes questions des paramètres étudiés.

  20. Premiere of TV Documentary Choe Chi-won Held in Beijing

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    <正>The premiere of the TV documentary Choe Chi-won was held in Beijing on August 19. The TV documentary,a joint production by the CPAFFC,the Jiangsu Provincial Association for Cultural Exchange with Foreign Countries,the

  1. Reactor Neutrinos

    OpenAIRE

    Soo-Bong Kim; Thierry Lasserre; Yifang Wang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  2. BOILING REACTORS

    Science.gov (United States)

    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)

  3. Experimental Plan and Irradiation Target Design for FeCrAl Embrittlement Screening Tests Conducted Using the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-06-26

    The objective of the FeCrAl embrittlement screening tests being conducted through the use of Oak Ridge National Laboratories (ORNL) High Flux Isotope Reactor is to provide data on the radiation-induced changes in the mechanical properties including radiation-induced hardening and embrittlement through systematic testing and analysis. Data developed on the mechanical properties will be supported by extensive microstructural evaluations to assist in the development of structure-property relationships and provide a sound, fundamental understanding of the performance of FeCrAl alloys in intense neutron radiation fields. Data and analysis developed as part of this effort will be used to assist in the determination of FeCrAl alloys as a viable material for commercial light water reactor (LWR) applications with a primary focus as an accident tolerant cladding.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-09-15

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

  5. PRELIMINARY RESULTS OF THE AGC-4 IRRADIATION IN THE ADVANCED TEST REACTOR AND DESIGN OF AGC-5 (HTR16-18469)

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gas Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results

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

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, L.R.; Wittman, R. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Pierson, B.P. [Univ. of Michigan, Ann Arbor, MI 48109 (United States); Metz, L.A.; Payne, R.; Finn, E.C.; Friese, J.I. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States)

    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)

  7. Membrane reactor. Membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shindo, Y.; Wakabayashi, K. (National Chemical Laboratory for Industry, Tsukuba (Japan))

    1990-08-05

    Many reaction examples were introduced of membrane reactor, to be on the point of forming a new region in the field of chemical technology. It is a reactor to exhibit excellent function, by its being installed with membrane therein, and is generally classified into catalyst function type and reaction promotion type. What firstly belongs to the former is stabilized zirconia, where oxygen, supplied to the cathodic side of membrane with voltage, impressed thereon, becomes O {sup 2 {minus}} to be diffused through the membrane and supplied, as variously activated oxygenous species, on the anodic side. Examples with many advantages can be given such as methane coupling, propylene oxidation, methanating reaction of carbon dioxide, etc. Apart, palladium film and naphion film also belong to the former. While examples of the latter comprise, among others, decomposition of hydrogen sulfide by porous glass film and dehydrogenation of cyclohexane or palladium alloy film, which are expected to be developed and materialized in the industry. 33 refs., 8 figs.

  8. The feasibility of small size specimens for testing of environmentally assisted cracking of irradiated materials and of materials under irradiation in reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Toivonen, A.; Moilanen, P.; Pyykkoenen, M.; Taehtinen, S.; Rintamaa, R.; Saario, T. [Valtion Teknillinen Tutkimuskeskus, Espoo (Finland)

    1998-11-01

    Environmentally assisted cracking (EAC) of core materials has become an increasingly important issue of downtime and maintenance costs in nuclear power plants. Small size specimens are necessary in stress corrosion testing of irradiated materials because of difficulties in handling high dose rate materials and because of restricted availability of the materials. The drawback of using small size specimens is that in some cases they do not fulfil the requirements of the relevant testing standards. Recently VTT has developed J-R testing with irradiated and non-irradiated sub size 3 PB specimens, both in inert and in LWR environments. Also, a new materials testing system which will enable simultaneous multiple specimen testing both in laboratory conditions and in operating reactor core is under development. The new testing system will utilize Charpy and sub size 3 PB specimens. The feasibility study of the system has been carried out using different materials. Fracture resistance curves of a Cu-Zr-Cr alloy are shown to be independent of the specimen geometry and size, to some extent. Results gained from tests in simulated boiling water reactor (BWR) water are presented for sensitized SIS 2333 stainless steel. The experimental results indicate that the size of the plastic zone or stress triaxiality must be further studied although no significant effect on the environmentally assisted crack growth rate was observed. (orig.)

  9. Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kontogeorgakos, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Derstine, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Bauer, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Stevens, J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2013-06-01

    The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO2 particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.

  10. JHR Project: a future Material Testing Reactor working as an International user Facility: The key-role of instrumentation in support to the development of modern experimental capacity

    Energy Technology Data Exchange (ETDEWEB)

    Bignan, G. [CEA, DEN, DER, JHR user Facility Interface Manager' , Cadarache, F-13108 St-Paul-Lez-Durance (France); Gonnier, C. [CEA, DEN, DER, SRJH Jules Horowitz Reactor Service, Cadarache, F-13108 St-Paul-Lez-Durance (France); Lyoussi, A.; Villard, J.F.; Destouches, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Chauvin, J.P. [CEA,DEN, DER, SPEX, Experimental Physics Service, Cadarache, F-13108 St-Paul-Lez-Durance (France); Maugard, B. [CEA, DEN, DER, Reactor Department Studies, Cadarache, F-13108 St-Paul-Lez-Durance (France)

    2015-07-01

    Research and development on fuel and material behaviour under irradiation is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. These needs mainly deal with a constant improvement of performances and safety in order to optimize the fuel cycle and hence to reach nuclear energy sustainable objectives. A sustainable nuclear energy requires a high level of performances in order to meet specific needs such as: - Pursuing improvement of the performances and safety of present and coming water cooled reactor technologies. This will require a continuous R and D support following a long-term trend driven by the plant life management, safety demonstration, flexibility and economics improvement. Experimental irradiations of structure materials are necessary to anticipate these material behaviours and will contribute to their optimisation. - Upgrading continuously nuclear fuel technology in present and future nuclear power plants to achieve better performances and to optimise the fuel cycle keeping the best level of safety. Fuel evolution for generation II, III and III+ is a key stake requiring developments, qualification tests and safety experiments to ensure the competitiveness and safety: experimental tests exploring the full range of fuel behaviour determine fuel stability limits and safety margins, as a major input for the fuel reliability analysis. To perform such accurate and innovative progress and developments, specific and ad hoc instrumentation, irradiation devices, measurement methods are necessary to be set up inside or beside the material testing reactor (MTR) core. These experiments require beforehand in situ and on line sophisticated measurements to accurately determine different key parameters such as thermal and fast neutron fluxes and nuclear heating in order to precisely monitor and control the conducted assays. The new Material Testing Reactor JHR (Jules Horowitz Reactor) currently under

  11. On the Safety and Performance Demonstration Tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and Validation and Verification of Computational Codes

    Directory of Open Access Journals (Sweden)

    Jong-Bum Kim

    2016-10-01

    Full Text Available The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR has been developed and the validation and verification (V&V activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1, produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  12. High-intensity running in English FA Premier League soccer matches

    DEFF Research Database (Denmark)

    Bradley, Paul S.; Sheldon, William; Wooster, Blake

    2009-01-01

    The aims of this study were to (1) determine the activity profiles of a large sample of English FA Premier League soccer players and (2) examine high-intensity running during elite-standard soccer matches for players in various playing positions. Twenty-eight English FA Premier League games were.......01), attackers (2341 m, s=575, P game, high-intensity running distance was approximately 20% less than in the first 15-min period for wide midfielders (467 m, s=104 vs. 589 m, s=134, P ....01) and without ball possession (229 m, s=85 vs. 278 m, s=97, P game. Mean recovery time between very high-intensity running bouts was 72 s (s=28), with a 28% longer recovery time during the last 15 min than the first 15 min of the game (83 s, s=26 vs...

  13. Black Generation Y gender differences in Premier Soccer League spectator motives : sport marketing

    OpenAIRE

    T.E. Mofokeng; Bevan-Dye, A.L.

    2014-01-01

    The purpose of this study was to determine whether there are gender differences concerning Premier Soccer League (PSL) spectator motives amongst black Generation Y students in South Africa. In South Africa, the black Generation Y cohort (individuals born between 1986 and 2005) represents an important but under-researched market segment in that, in 2013, they made up 32 percent of the country's population. From a PSL marketing perspective, understanding the motives that drive game spectatorshi...

  14. Market segmentation in two-sided markets : tv rights for premier league

    OpenAIRE

    Kind, Hans Jarle; Sørgard, Lars

    2012-01-01

    This paper analyzes market segmentation in a two-sided market that consists of media consumers and advertisers. The analysis is motivated by a European Court of Justice Decision in October 2011, which allowed viewers to take advantage of international price differences and buy access to Premier League TV matches from whichever country they like. We compare complete market segmentation with the new situation where consumers can purchase from abroad (allowing for passive sales). Clearly, such a...

  15. ABOUT THE SMART SPORTS DEVELOPMENT. EVIDENCE FROM THE UK PREMIERE LEAGUE

    OpenAIRE

    Vlad Ionut Dumitrache

    2016-01-01

    Smart economy implies the development of key factors like global economy growth, competition, economic progress, economic prosperity, innovation. In the European top-level football, like the case of the British Premier League, financial indicators have demonstrated that the factors that define smart economy can be identified. The new rules of the financial fair-play policies and the ever growing revenues for television rights have created a new market in sports economy, one that identifies it...

  16. Prediction calculations and experiments for the first criticality of the 10 MW High Temperature Gas-cooled Reactor-Test Module

    Energy Technology Data Exchange (ETDEWEB)

    Jing Xingqing E-mail: jingxq@d103.inet.tsinghua.edu.cn; Xu Xiaolin; Yang Yongwei; Qu Ronghong

    2002-10-01

    The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is a pebble bed experimental reactor built by the Institute of Nuclear Energy Technology (INET), Tsinghua University. This paper introduces the first critical prediction calculations and the experiments for the HTR-10. The German VSOP neutronics code is used for the prediction calculations of the first loading. The characteristics of pebble-bed high temperature gas-cooled reactors are taken into account, including the double heterogeneity of the fuel element, the buckling feedback of the spectrum calculation, the effect of the mixture of fuel elements and graphite balls, and the correction of the diffusion coefficients in the upper cavity based on transport theory. Also considered are the effects of impurities in the fuel elements, in the graphite balls and in the reflector graphite on the reactivity. The number of fuel elements and graphite balls in the initial core is predicted to provide reference for the first criticality experiment. The critical experiment adopts a method of extrapolating to approach criticality. The first criticality was attained on December 1, 2000. The first criticality experiment shows that the predicted critical number of the fuel elements and graphite balls is in close agreement with the experimental results. Their relative error is less than 1.0%, implying the physical predictions and the results of the criticality experiment are much beyond expectations.

  17. Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanan, N. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-04-07

    The document compiles in a single volume several verification and validation works done for the PLTEMP/ANL code during the years of its development and improvement. Some works that are available in the open literature are simply referenced at the outset, and are not included in the document. PLTEMP has been used in conversion safety analysis reports of several US and foreign research reactors that have been licensed and converted. A list of such reactors is given. Each chapter of the document deals with the verification or validation of a specific model. The model verification is usually done by comparing the code with hand calculation, Microsoft spreadsheet calculation, or Mathematica calculation. The model validation is done by comparing the code with experimental data or a more validated code like the RELAP5 code.

  18. Logistique de transport pour le projet LHC enseignements des premiers secteurs

    CERN Document Server

    Prodon, S

    2003-01-01

    Ce papier dresse un premier bilan de la logistique de transport mise en place pour l'installation du LHC. Les moyens de planification mis en oeuvre seront tout d'abord évoqués avec notamment les réunions avec les groupes utilisateurs, l'élaboration de procédures de transport, la génération de listings d'articles à transporter ou encore l'établissement d'un planning des ressources. Cependant, les premiers travaux d'installation du LHC ont fait apparaître des divergences importantes entre le planning logistique établi et la réalité du terrain. Ces écarts seront analysés, qu'il s'agisse de différences sur le volume de matériel à acheminer, d'opérations non planifiées, de changements de plannings entraînant de longues et délicates traversées de chantiers ou de manque de planification des besoins en personnel dans certaines zones. Tous ces enseignements acquis au cours des premiers travaux devraient permettre de dégager des voies d'amélioration à mettre en place pour les prochains secteur...

  19. Technical Letter Report, An Evaluation of Ultrasonic Phased Array Testing for Reactor Piping System Components Containing Dissimilar Metal Welds, JCN N6398, Task 2A

    Energy Technology Data Exchange (ETDEWEB)

    Diaz, Aaron A.; Cinson, Anthony D.; Crawford, Susan L.; Anderson, Michael T.

    2009-11-30

    Research is being conducted for the U.S. Nuclear Regulatory Commission at the Pacific Northwest National Laboratory to assess the effectiveness and reliability of advanced nondestructive examination (NDE) methods for the inspection of light-water reactor components. The scope of this research encom¬passes primary system pressure boundary materials including dissimilar metal welds (DMWs), cast austenitic stainless steels (CASS), piping with corrosion-resistant cladding, weld overlays, inlays and onlays, and far-side examinations of austenitic piping welds. A primary objective of this work is to evaluate various NDE methods to assess their ability to detect, localize, and size cracks in steel components that challenge standard and/or conventional inspection methodologies. This interim technical letter report provides a summary of a technical evaluation aimed at assessing the capabilities of phased-array (PA) ultrasonic testing (UT) methods as applied to the inspection of small-bore DMW components that exist in the reactor coolant systems (RCS) of pressurized water reactors (PWRs). Operating experience and events such as the circumferential cracking in the reactor vessel nozzle-to-RCS hot leg pipe at V.C. Summer nuclear power station, identified in 2000, show that in PWRs where primary coolant water (or steam) are present under normal operation, Alloy 82/182 materials are susceptible to pressur