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Sample records for high flux engineering test reactor

  1. Feature of high flux engineering test reactor and its role in nuclear power development

    SciTech Connect

    Guangquan, L.

    1988-01-01

    The High Flux Engineering Test Reactor (HFETR) designed and built by China own efforts reached to its initial criticality on Dec. 27, 1979, and then achieved high power operation on Dec. 16, 1980. Until Nov. 11, 1986, the reactor had been operated for thirteen cycles. The paper presents briefly main feature of HFETR and its utilization during past years. The paper also deals with its role in nuclear power development. Finally, author gives his opinion on comprehensive utilization of HFETR.

  2. High flux reactor

    DOEpatents

    Lake, James A.; Heath, Russell L.; Liebenthal, John L.; DeBoisblanc, Deslonde R.; Leyse, Carl F.; Parsons, Kent; Ryskamp, John M.; Wadkins, Robert P.; Harker, Yale D.; Fillmore, Gary N.; Oh, Chang H.

    1988-01-01

    A high flux reactor is comprised of a core which is divided into two symetric segments housed in a pressure vessel. The core segments include at least one radial fuel plate. The spacing between the plates functions as a coolant flow channel. The core segments are spaced axially apart such that a coolant mixing plenum is formed between them. A channel is provided such that a portion of the coolant bypasses the first core section and goes directly into the mixing plenum. The outlet coolant from the first core segment is mixed with the bypass coolant resulting in a lower inlet temperature to the lower core segment.

  3. ENGINEERING TEST REACTOR

    DOEpatents

    De Boisblanc, D.R.; Thomas, M.E.; Jones, R.M.; Hanson, G.H.

    1958-10-21

    Heterogeneous reactors of the type which is both cooled and moderated by the same fluid, preferably water, and employs highly enriched fuel are reported. In this design, an inner pressure vessel is located within a main outer pressure vessel. The reactor core and its surrounding reflector are disposed in the inner pressure vessel which in turn is surrounded by a thermal shield, Coolant fluid enters the main pressure vessel, fiows downward into the inner vessel where it passes through the core containing tbe fissionable fuel assemblies and control rods, through the reflector, thence out through the bottom of the inner vessel and up past the thermal shield to the discharge port in the main vessel. The fuel assemblles are arranged in the core in the form of a cross having an opening extending therethrough to serve as a high fast flux test facility.

  4. Graphite irradiation testing for HTR technology at the High Flux Reactor in Petten

    NASA Astrophysics Data System (ADS)

    Vreeling, J. A.; Wouters, O.; Laan, J. G. van der

    2008-10-01

    In 2001 the Nuclear Research and Consultancy Group started a large graphite irradiation program for the development of High Temperature Reactor technology in the European framework. The irradiation experiments, containing present day available graphite grades, are performed at the High Flux Reactor in Petten. The grades are NBG-10, NBG-17, NBG-18, NBG-20, NBG-25, PCEA, PPEA, PCIB, LPEB, IG-110 and IG-430. In the fifth framework programme (2001-2004) and sixth framework programme (2005-2009) four irradiation experiments are foreseen, resulting in design curves at irradiation temperatures between 650 °C and 950 °C. The post-irradiation testing is focused on dimensional changes, dynamic Young's modulus, coefficient of thermal expansion and coefficient of thermal conductivity. The irradiation programme and preliminary results from the first irradiation experiment at 750 °C to 8 dpa will be discussed in this paper.

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

    SciTech Connect

    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.

  6. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

    NASA Astrophysics Data System (ADS)

    Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Bréaud, S.; Oriol, L.; Villard, J.-F.

    2011-03-01

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

  7. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

    SciTech Connect

    Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F.; Lopez, A. Legrand

    2011-03-15

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 x 10{sup 20} n/cm{sup 2}. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

  8. New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions.

    PubMed

    Geslot, B; Vermeeren, L; Filliatre, P; Lopez, A Legrand; Barbot, L; Jammes, C; Bréaud, S; Oriol, L; Villard, J-F

    2011-03-01

    Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 10(20) n∕cm(2). A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

  9. British high flux beam reactor.

    PubMed

    Egelstaff, P A

    1970-10-24

    The neutron scattering technique has become an accepted method for the study of condensed matter. Because of the great scientific and technical value of neutron experiments and the growing body of users, several proposals have been made during the past decade for a nuclear reactor devoted primarily to this technique. This article reviews the reasons for and history behind these proposals.

  10. Seismic study of high-temperature engineering test reactor core graphite structures

    SciTech Connect

    Iyoku, T.; Inagaki, Y.; Shiozawa, S. . Oarai Research Establishment); Nishiguchi, I. )

    1992-08-01

    This paper discusses the High-Temperature Engineering Test Reactor (HTTR) a 30-MW (thermal) helium gas-cooled reactor with a core composed of prismatic graphite blocks piled on core support structures. Safety analyses have been made for the seismic design of the HTTR core using a two-dimensional seismic analysis code called SONATINA-2V, which was developed by the Japan Atomic Energy Research Institute. To evaluate the validity of the SONATINA-2V code and confirm the structural integrity of the core graphite blocks, large-scale seismic tests are conducted using a half-scale vertical section model and a full-scale seven-column model of the core graphite blocks and the core support structures. The test results are in good agreement with the analytical ones, and the validity of the analysis code is confirmed. The structural integrity of the core graphite blocks is confirmed by both analytical and test results.

  11. A unique high heat flux facility for testing hypersonic engine components

    NASA Technical Reports Server (NTRS)

    Melis, Matthew E.; Gladden, Herbert J.

    1990-01-01

    This paper describes the Hot Gas Facility, a unique, reliable, and cost-effective high-heat-flux facility for testing hypersonic engine components developed at the NASA Lewis Research Center. The Hot Gas Facility is capable of providing heat fluxes ranging from 200 Btu/sq ft per sec on flat surfaces up to 8000 Btu/sq ft per sec at a leading edge stagnation point. The usefulness of the Hot Gas Facility for the NASP community was demonstrated by testing hydrogen-cooled structures over a range of temperatures and pressures. Ranges of the Reynolds numbers, Prandtl numbers, enthalpy, and heat fluxes similar to those expected during hypersonic flights were achieved.

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

    SciTech Connect

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

  14. A unique high heat flux facility for testing hypersonic engine components

    NASA Technical Reports Server (NTRS)

    Melis, Matthew E.; Gladden, Herbert J.

    1990-01-01

    A major concern in advancing the state-of-the-art technologies for hypersonic vehicles is the development of an aeropropulsion system capable of withstanding high thermal loads expected during hypersonic flights. Consequently, there is a need for experimental facilities capable of providing a high heat flux environment for testing compound concepts and verifying analyses. A hydrogen/oxygen rocket engine was developed to provide a high enthalpy/high heat flux environment for component evaluation. This Hot Gas Facility is capable of providing heat fluxes ranging from 200 (on flat surfaces) up to 8000 Btu per sq ft per sec (at a leading edge stagnation point). Gas temperatures up to 5500 R can be attained as well as Reynolds numbers up to 360,000 per ft. Test articles such as cowl leading edges, transpiration-cooled seals, fuel injectors, and cooled panel concepts can be evaluated with gaseous hydrogen as coolant. This facility and its configuration and test capabilities are discussed. Results from flow characterization experiments are also shown and their implications considered.

  15. HFBR handbook, 1992: High flux beam reactor

    SciTech Connect

    Axe, J.D.; Greenberg, R.

    1992-10-01

    Welcome to the High Flux Beam Reactor (HFBR), one of the world premier neutron research facilities. This manual is intended primarily to acquaint outside users (and new Brookhaven staff members) with (almost) everything they need to know to work at the HFBR and to help make the stay at Brookhaven pleasant as well as profitable. Safety Training Programs to comply with US Department of Energy (DOE) mandates are in progress at BNL. There are several safety training requirements which must be met before users can obtain unescorted access to the HFBR. The Reactor Division has prepared specific safety training manuals which are to be sent to experimenters well in advance of their expected arrival at BNL to conduct experiments. Please familiarize yourself with this material and carefully pay strict attention to all the safety and security procedures that are in force at the HFBR. Not only your safety, but the continued operation of the facility, depends upon compliance.

  16. System Engineering Program Applicability for the High Temperature Gas-Cooled Reactor (HTGR) Component Test Capability (CTC)

    SciTech Connect

    Jeffrey Bryan

    2009-06-01

    This white paper identifies where the technical management and systems engineering processes and activities to be used in establishing the High Temperature Gas-cooled Reactor (HTGR) Component Test Capability (CTC) should be addressed and presents specific considerations for these activities under each CTC alternative

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

    NASA Astrophysics Data System (ADS)

    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.

  18. Evaluation of the Start-Up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Annular Core Loadings)

    SciTech Connect

    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

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

    SciTech Connect

    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.

  20. High Flux Metallic Membranes for Hydrogen Recovery and Membrane Reactors

    SciTech Connect

    Buxbaum, Robert

    2010-06-30

    We made and tested over 250 new alloys for use as lower cost, higher flux hydrogen extraction membrane materials. Most of these were intermetallic, or contained significant intermetallic content, particularly based on B2 alloy compositions with at least one refractory component; B2 intermetallics resemble BCC alloys, in structure, but the atoms have relatively fixed positions, with one atom at the corners of the cube, the other at the centers. The target materals we were looking for would contain little or no expensive elements, no strongly toxic or radioactive elements, would have high flux to hydrogen, while being fabricable, brazable, and relatively immune to hydrogen embrittlement and corrosion in operation. The best combination of properties of the membrane materials we developed was, in my opinion, a Pd-coated membrane consisting of V -9 atomic % Pd. This material was relatively cheap, had 5 times the flux of Pd under the same pressure differential, was reasonably easy to fabricate and braze, and not bad in terms of embrittlement. Based on all these factors we project, about 1/3 the cost of Pd, on an area basis for a membrane designed to last 20 years, or 1/15 the cost on a flux basis. Alternatives to this membrane replaced significant fractions of the Pd with Ni and or Co. The cost for these membranes was lower, but so was the flux. We produced successful brazed products from the membrane materials, and made them into flat sheets. We tested, unsuccessfully, several means of fabricating thematerials into tubes, and eventually built a membrane reactor using a new, flat-plate design: a disc and doughnut arrangement, a design that seems well- suited to clean hydrogen production from coal. The membranes and reactor were tested successfully at Western Research. A larger equipment company (Chart Industries) produced similar results using a different flat-plate reactor design. Cost projections of the membrane are shown to be attractive.

  1. Reactor operations: Brookhaven Medical Research Reactor, Brookhaven High Flux Beam Reactor. Informal report, June 1995

    SciTech Connect

    1995-06-01

    Part one of this report gives the operating history of the Brookhaven Medical Research Reactor for the month of June. Also included are the BMRR technical safety surveillance requirements record and the summary of BMRR irradiations for the month. Part two gives the operating histories of the Brookhaven High Flux Beam Reactor and the Cold Neutron Facility at HFBR for June. Also included are the HFBR technical safety surveillance requirements record and the summary of HFBR irradiations for the month.

  2. Experimental Plan and Irradiation Target Design for FeCrAl Embrittlement Screening Tests Conducted Using the High Flux Isotope Reactor

    SciTech Connect

    Field, Kevin G.; Howard, Richard H.; Yamamoto, Yukinori

    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.

  3. Design and Nuclear-Safety Related Simulations of Bare-Pellet Test Irradiations for the Production of Pu-238 in the High Flux Isotope Reactor using COMSOL

    SciTech Connect

    Freels, James D; Jain, Prashant K; Hobbs, Randy W

    2012-01-01

    The Oak Ridge National Laboratory (ORNL)is developing technology to produce plutonium-238 for the National Aeronautics and Space Administration (NASA) as a power source material for powering vehicles while in deep-space[1]. The High Flux Isotope Reactor (HFIR) of ORNL has been utilized to perform test irradiations of incapsulated neptunium oxide (NpO2) and aluminum powder bare pellets for purposes of understanding the performance of the pellets during irradiation[2]. Post irradiation examinations (PIE) are currently underway to assess the effect of temperature, thermal expansion, swelling due to gas production, fission products, and other phenomena

  4. Transmutation of MA in the high flux thermal reactor

    NASA Astrophysics Data System (ADS)

    Liu, Bin; Hu, Wenchao; Wang, Kai; Huang, Liming; Ouyang, Xiaoping; Tu, Jing; Zhu, Yangni

    2013-06-01

    We study the MA transmutation characteristics in the high flux thermal reactor, our calculation shows that different MA nuclides may have the drastically different effects on keff, Np-237, Am-241 and Am-243 decrease keff greatly, Cm-244 affects the keff slightly, but Cm-245 boosts the keff significantly. The MA nuclides actually can act as the burnable poisons in the thermal reactors. The SCALE simulation shows that after 300-day-exposure in high flux thermal reactor the disappearance rate of Np-237, Am-241 and Am-243 are 73.7%, 98.1% and 82.8% respectively. The SCALE simulation results also show that the total transmutation rate of MA nuclides by fission which include direct and indirect fission after 300-day-exposure in the high flux thermal reactor is 6.3%. The SCALE simulation indicates at least 44.2% MA nuclides transmute to plutonium isotopes by various reactions and 63% of the Pu-238 in the MOX fuel is consumed during 300-day-exposure in this reactor.

  5. Jules Horowitz Reactor: a high performance material testing reactor

    NASA Astrophysics Data System (ADS)

    Iracane, Daniel; Chaix, Pascal; Alamo, Ana

    2008-04-01

    The physical modelling of materials' behaviour under severe conditions is an indispensable element for developing future fission and fusion systems: screening, design, optimisation, processing, licensing, and lifetime assessment of a new generation of structure materials and fuels, which will withstand high fast neutron flux at high in-service temperatures with the production of elements like helium and hydrogen. JANNUS and other analytical experimental tools are developed for this objective. However, a purely analytical approach is not sufficient: there is a need for flexible experiments integrating higher scales and coupled phenomena and offering high quality measurements; these experiments are performed in material testing reactors (MTR). Moreover, complementary representative experiments are usually performed in prototypes or dedicated facilities such as IFMIF for fusion. Only such a consistent set of tools operating on a wide range of scales, can provide an actual prediction capability. A program such as the development of silicon carbide composites (600-1200 °C) illustrates this multiscale strategy. Facing the long term needs of experimental irradiations and the ageing of present MTRs, it was thought necessary to implement a new generation high performance MTR in Europe for supporting existing and future nuclear reactors. The Jules Horowitz Reactor (JHR) project copes with this context. It is funded by an international consortium and will start operation in 2014. JHR will provide improved performances such as high neutron flux ( 10 n/cm/s above 0.1 MeV) in representative environments (coolant, pressure, temperature) with online monitoring of experimental parameters (including stress and strain control). Experimental devices designing, such as high dpa and small thermal gradients experiments, is now a key objective requiring a broad collaboration to put together present scientific state of art, end-users requirements and advanced instrumentation. To cite this

  6. Decommissioning of the high flux beam reactor at Brookhaven Lab

    SciTech Connect

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

    2011-07-01

    The high-flux beam reactor (HFBR) at the Brookhaven National Laboratory was a heavy water cooled and moderated reactor that achieved criticality on Oct. 31, 1965. It operated at a power level of 40 megawatts. An equipment upgrade in 1982 allowed operations at 60 megawatts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 megawatts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of groundwater from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost three years for safety and environmental reviews. In November 1999 the United States Dept. of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel, is presently under 24/7 surveillance for safety. Detailed dosimetry performed for the HFBR decommissioning during 1996-2009 is described in the paper. (authors)

  7. High Flux Isotope Reactor system RELAP5 input model

    SciTech Connect

    Morris, D.G.; Wendel, M.W.

    1993-01-01

    A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model.

  8. Decommissioning of the High Flux Beam Reactor at Brookhaven Lab

    SciTech Connect

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

    2011-05-27

    The High Flux Beam Reactor at the Brookhaven National Laboratory was a heavy water cooled and moderated reactor that achieved criticality on October 31, 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shutdown in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor’s spent fuel pool. The reactor remained shutdown for almost three years for safety and environmental reviews. In November 1999 the United States Department of Energy decided to permanently shutdown the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR cleanup conducted during 1999-2009 will be described in the paper.

  9. Performance and safety parameters for the high flux isotope reactor

    SciTech Connect

    Ilas, G.; Primm III, T.

    2012-07-01

    A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDF/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data. (authors)

  10. Flux trap void effects in the high-flux isotope reactor with curium targets

    SciTech Connect

    Rothrock, R.B.

    1990-06-01

    The reactivity effect of voids in the high-flux isotope reactor (HFIR) flux trap/target region has been reevaluated to determine the impact of the use of curium target material in lieu of the {sup 242}Pu targets originally used. The HFIR, located at the Oak Ridge National Laboratory (ORNL) and operated for the US Department of Energy by Martin Marietta Energy Systems, Inc., produces transplutonium elements for research and industrial purposes by irradiating target materials in a central flux trap region of intense thermal neutron flux. The target region void worth was extensively investigated during the HFIR nuclear design and critical experiments. Subsequently, as larger quantities of heavy elements became available, HFIR targets were fabricated from curium. This change substantially increased the amount of transplutonium isotopes present in the target during irradiation; it is therefore necessary to determine the impact on the target region void worth. Nuclear calculations of the target void worth were made using the two-dimensional transport code DORT. To benchmark the methods and data, flux trap void worth experiments performed during the HFIR start-up physics tests were calculated with the two-dimensional model for void worth measurements with and without the simulated target present in the flux trap.

  11. Component and system simulation models for High Flux Isotope Reactor

    SciTech Connect

    Sozer, A.

    1989-08-01

    Component models for the High Flux Isotope Reactor (HFIR) have been developed. The models are HFIR core, heat exchangers, pressurizer pumps, circulation pumps, letdown valves, primary head tank, generic transport delay (pipes), system pressure, loop pressure-flow balance, and decay heat. The models were written in FORTRAN and can be run on different computers, including IBM PCs, as they do not use any specific simulation languages such as ACSL or CSMP. 14 refs., 13 figs.

  12. Neutronics Modeling of the High Flux Isotope Reactor using COMSOL

    SciTech Connect

    Chandler, David; Primm, Trent; Freels, James D; Maldonado, G Ivan

    2011-01-01

    The High Flux Isotope Reactor located at the Oak Ridge National Laboratory is a versatile 85 MWth research reactor with cold and thermal neutron scattering, materials irradiation, isotope production, and neutron activation analysis capabilities. HFIR staff members are currently in the process of updating the thermal hydraulic and reactor transient modeling methodologies. COMSOL Multiphysics has been adopted for the thermal hydraulic analyses and has proven to be a powerful finite-element-based simulation tool for solving multiple physics-based systems of partial and ordinary differential equations. Modeling reactor transients is a challenging task because of the coupling of neutronics, heat transfer, and hydrodynamics. This paper presents a preliminary COMSOL-based neutronics study performed by creating a two-dimensional, two-group, diffusion neutronics model of HFIR to study the spatially-dependent, beginning-of-cycle fast and thermal neutron fluxes. The 238-group ENDF/B-VII neutron cross section library and NEWT, a two-dimensional, discrete-ordinates neutron transport code within the SCALE 6 code package, were used to calculate the two-group neutron cross sections required to solve the diffusion equations. The two-group diffusion equations were implemented in the COMSOL coefficient form PDE application mode and were solved via eigenvalue analysis using a direct (PARDISO) linear system solver. A COMSOL-provided adaptive mesh refinement algorithm was used to increase the number of elements in areas of largest numerical error to increase the accuracy of the solution. The flux distributions calculated by means of COMSOL/SCALE compare well with those calculated with benchmarked three-dimensional MCNP and KENO models, a necessary first step along the path to implementing two- and three-dimensional models of HFIR in COMSOL for the purpose of studying the spatial dependence of transient-induced behavior in the reactor core.

  13. Conversion feasibility studies for the Grenoble high flux reactor

    SciTech Connect

    Mo, S.C.; Matos, J.E.

    1989-01-01

    Feasibility studies for conversion of the High Flux Reactor (RHF) at Grenoble France have been performed at the Argonne National Laboratory in cooperation with the Institut Laue-Langevin (ILL). The uranium densities required for conversion of the RHF to reduced enrichment fuels were computed to be 7.9 g/cm{sup 3} with 20% enrichment, 4.8 g/cm{sup 3} with 29% enrichment, and 2.8 g/cm{sup 3} with 45% enrichment. Thermal flux reductions at the peak in the heavy water reflector were computed to be 3% with 45% enriched fuel and 7% with 20% enriched fuel. In each case, the reactor's 44 day cycle length was preserved and no changes were made in the fuel element geometry. If the cladding thickness could be reduced from 0.38 mm to 0.30 mm, the required uranium density with 20% enrichment would be about 6.0 g/cm{sup 3} and the thermal flux reduction at the peak in the heavy water reflector would be about 7%. Significantly higher uranium densities are required in the RHF than in heavy water reactors with more conventional designs because the neutron spectrum is much harder in the RHF. Reduced enrichment fuels with the uranium densities required for use in the RHF are either not available or are not licensable at the present time. 6 refs., 6 figs., 3 tabs.

  14. High Flux Isotope Reactor cold neutron source reference design concept

    SciTech Connect

    Selby, D.L.; Lucas, A.T.; Hyman, C.R.

    1998-05-01

    In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

  15. Rebuilding the Brookhaven high flux beam reactor: A feasibility study

    SciTech Connect

    Brynda, W.J.; Passell, L.; Rorer, D.C.

    1995-01-01

    After nearly thirty years of operation, Brookhaven`s High Flux Beam Reactor (HFBR) is still one of the world`s premier steady-state neutron sources. A major center for condensed matter studies, it currently supports fifteen separate beamlines conducting research in fields as diverse as crystallography, solid-state, nuclear and surface physics, polymer physics and structural biology and will very likely be able to do so for perhaps another decade. But beyond that point the HFBR will be running on borrowed time. Unless appropriate remedial action is taken, progressive radiation-induced embrittlement problems will eventually shut it down. Recognizing the HFBR`s value as a national scientific resource, members of the Laboratory`s scientific and reactor operations staffs began earlier this year to consider what could be done both to extend its useful life and to assure that it continues to provide state-of-the-art research facilities for the scientific community. This report summarizes the findings of that study. It addresses two basic issues: (i) identification and replacement of lifetime-limiting components and (ii) modifications and additions that could expand and enhance the reactor`s research capabilities.

  16. Emergency Procedure Training for Reactor Operators at the High Flux Beam Reactor for Brookhaven National Laboratory.

    ERIC Educational Resources Information Center

    Reyer, Ronald

    A project was conducted to analyze, design, develop, implement, and evaluate an instructional unit intended to improve the diagnostic skills of operating personnel in responding to abnormal and emergency conditions at the High Flux Beam Reactor at Brookhaven National Laboratory. Research was conducted on the occurrence of emergencies at similar…

  17. Fabrication of control rods for the High Flux Isotope Reactor

    SciTech Connect

    Sease, J.D.

    1998-03-01

    The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A.

  18. High flux isotope reactor cold source preconceptual design study report

    SciTech Connect

    Selby, D.L.; Bucholz, J.A.; Burnette, S.E.

    1995-12-01

    In February 1995, the deputy director of Oak Ridge National Laboratory (ORNL) formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced Neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. The anticipated cold source will consist of a cryogenic LH{sub 2} moderator plug, a cryogenic pump system, a refrigerator that uses helium gas as a refrigerant, a heat exchanger to interface the refrigerant with the hydrogen loop, liquid hydrogen transfer lines, a gas handling system that includes vacuum lines, and an instrumentation and control system to provide constant system status monitoring and to maintain system stability. The scope of this project includes the development, design, safety analysis, procurement/fabrication, testing, and installation of all of the components necessary to produce a working cold source within an existing HFIR beam tube. This project will also include those activities necessary to transport the cold neutron beam to the front face of the present HFIR beam room. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and research and development (R and D), (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the preconceptual phase and establishes the concept feasibility. The information presented includes the project scope, the preliminary design requirements, the preliminary cost and schedule, the preliminary performance data, and an outline of the various plans for completing the project.

  19. Next generation fuel irradiation capability in the High Flux Reactor Petten

    NASA Astrophysics Data System (ADS)

    Fütterer, Michael A.; D'Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco; Raison, Philippe; Bakker, Klaas; de Groot, Sander; Klaassen, Frodo

    2009-07-01

    This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

  20. Delivery of completed irradiation vehicles and the quality assurance document to the High Flux Isotope Reactor for irradiation

    SciTech Connect

    Petrie, Christian M.; McDuffee, Joel Lee; Katoh, Yutai; Terrani, Kurt A.

    2015-10-01

    This report details the initial fabrication and delivery of two Fuel Cycle Research and Development (FCRD) irradiation capsules (ATFSC01 and ATFSC02), with associated quality assurance documentation, to the High Flux Isotope Reactor (HFIR). The capsules and documentation were delivered by September 30, 2015, thus meeting the deadline for milestone M3FT-15OR0202268. These irradiation experiments are testing silicon carbide composite tubes in order to obtain experimental validation of thermo-mechanical models of stress states in SiC cladding irradiated under a prototypic high heat flux. This document contains a copy of the completed capsule fabrication request sheets, which detail all constituent components, pertinent drawings, etc., along with a detailed summary of the capsule assembly process performed by the Thermal Hydraulics and Irradiation Engineering Group (THIEG) in the Reactor and Nuclear Systems Division (RNSD). A complete fabrication package record is maintained by the THIEG and is available upon request.

  1. The High Flux Isotope Reactor (HFIR) cold source project at ORNL

    SciTech Connect

    Selby, D.L.; Lucas, A.T.; Chang, S.J.; Freels, J.D.

    1998-06-01

    Following the decision to cancel the Advanced Neutron Source (ANS) Project at Oak Ridge National Laboratory (ORNL), it was determined that a hydrogen cold source should be retrofitted into an existing beam tube of the High Flux Isotope Reactor (HFIR) at ORNL> The preliminary design of this system has been completed and an approval in principal of the design has been obtained from the internal ORNL safety review committees and the US Department of Energy (DOE) safety review committee. The cold source concept is basically a closed loop forced flow supercritical hydrogen system. The supercritical approach was chosen because of its enhanced stability in the proposed high heat flux regions. Neutron and gamma physics of the moderator have been analyzed using the 3D Monte Carlo code MCNP. A 3D structural analysis model of the moderator vessel, vacuum tube, and beam tube was completed to evaluate stress loadings and to examine the impact of hydrogen detonations in the beam tube. A detailed ATHENA system model of the hydrogen system has been developed to simulate loop performance under normal and off-normal transient conditions. Semi-prototypic hydrogen loop tests of the system have been performed at the Arnold Engineering Design Center (AEDC) located in Tullahoma, Tennessee to verify the design and benchmark the analytical system model. A 3.5 kW refrigerator system has been ordered and is expected to be delivered to ORNL by the end of this calendar year. The present schedule shows the assembling of the cold source loop on side during the fall of 1999 for final testing before insertion of the moderator plug assembly into the reactor beam tube during the end of the year 2000.

  2. Final report of the HFIR (High Flux Isotope Reactor) irradiation facilities improvement project

    SciTech Connect

    Montgomery, B.H.; Thoms, K.R.; West, C.D.

    1987-09-01

    The High-Flux Isotope Reactor (HFIR) has outstanding neutronics characteristics for materials irradiation, but some relatively minor aspects of its mechanical design severely limited its usefulness for that purpose. In particular, though the flux trap region in the center of the annular fuel elements has a very high neutron flux, it had no provision for instrumentation access to irradiation capsules. The irradiation positions in the beryllium reflector outside the fuel elements also have a high flux; however, although instrumented, they were too small and too few to replace the facilities of a materials testing reactor. To address these drawbacks, the HFIR Irradiation Facilities Improvement Project consisted of modifications to the reactor vessel cover, internal structures, and reflector. Two instrumented facilities were provided in the flux trap region, and the number of materials irradiation positions in the removable beryllium (RB) was increased from four to eight, each with almost twice the available experimental space of the previous ones. The instrumented target facilities were completed in August 1986, and the RB facilities were completed in June 1987.

  3. Advanced Multiphysics Thermal-Hydraulics Models for the High Flux Isotope Reactor

    SciTech Connect

    Jain, Prashant K; Freels, James D

    2015-01-01

    Engineering design studies to determine the feasibility of converting the High Flux Isotope Reactor (HFIR) from using highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL). This work is part of an effort sponsored by the US Department of Energy (DOE) Reactor Conversion Program. HFIR is a very high flux pressurized light-water-cooled and moderated flux-trap type research reactor. HFIR s current missions are to support neutron scattering experiments, isotope production, and materials irradiation, including neutron activation analysis. Advanced three-dimensional multiphysics models of HFIR fuel were developed in COMSOL software for safety basis (worst case) operating conditions. Several types of physics including multilayer heat conduction, conjugate heat transfer, turbulent flows (RANS model) and structural mechanics were combined and solved for HFIR s inner and outer fuel elements. Alternate design features of the new LEU fuel were evaluated using these multiphysics models. This work led to a new, preliminary reference LEU design that combines a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone. Preliminary results of estimated thermal safety margins are presented. Fuel design studies and model enhancement continue.

  4. Experimental Study of Thermal Crisis in Connection with Tokamak Reactor High Heat Flux Components

    SciTech Connect

    Gallo, D.; Giardina, M.; Castiglia, F.; Celata, G.P.; Mariani, A.; Zummo, G.; Cumo, M.

    2000-12-31

    The results of an experimental research on high heat flux thermal crisis in forced convective subcooled water flow, under operative conditions of interest to the thermal-hydraulic design of TOKAMAK fusion reactors, are here reported. These experiments, carried out in the framework of a collaboration between the Nuclear Engineering Department of Palermo University and the National Institute of Thermal - Fluid Dynamics of the ENEA - Casaccia (Rome), were performed on the STAF (Scambio Termico Alti Flussi) water loop and consisted, essentially, in a high speed photographic study which enabled focusing several information on bubble characteristics and flow patterns taking place during the burnout phenomenology.

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

    SciTech Connect

    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

  6. Level 1 Tornado PRA for the High Flux Beam Reactor

    SciTech Connect

    Bozoki, G.E.; Conrad, C.S.

    1994-05-01

    This report describes a risk analysis primarily directed at providing an estimate for the frequency of tornado induced damage to the core of the High Flux Beam Reactor (HFBR), and thus it constitutes a Level 1 Probabilistic Risk Assessment (PRA) covering tornado induced accident sequences. The basic methodology of the risk analysis was to develop a ``tornado specific`` plant logic model that integrates the internal random hardware failures with failures caused externally by the tornado strike and includes operator errors worsened by the tornado modified environment. The tornado hazard frequency, as well as earlier prepared structural and equipment fragility data, were used as input data to the model. To keep modeling/calculational complexity as simple as reasonable a ``bounding`` type, slightly conservative, approach was applied. By a thorough screening process a single dominant initiating event was selected as a representative initiator, defined as: ``Tornado Induced Loss of Offsite Power.`` The frequency of this initiator was determined to be 6.37E-5/year. The safety response of the HFBR facility resulted in a total Conditional Core Damage Probability of .621. Thus, the point estimate of the HFBR`s Tornado Induced Core Damage Frequency (CDF) was found to be: (CDF){sub Tornado} = 3.96E-5/year. This value represents only 7.8% of the internal CDF and thus is considered to be a small contribution to the overall facility risk expressed in terms of total Core Damage Frequency. In addition to providing the estimate of (CDF){sub Tornado}, the report documents, the relative importance of various tornado induced system, component, and operator failures that contribute most to (CDF){sub Tornado}.

  7. ENGINEERING TEST REACTOR, TRA642. CONTEXTUAL VIEW ORIENTATING ETR TO MTR. ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    ENGINEERING TEST REACTOR, TRA-642. CONTEXTUAL VIEW ORIENTATING ETR TO MTR. CAMERA IS ON ROOF OF MTR BUILDING AND FACES DUE SOUTH. MTR SERVICE BUILDING, TRA-635, IN LOWER RIGHT CORNER. STEEL FRAMES SHOW BUILDINGS TO BE ATTACHED TO ETR BUILDING. HIGH-BAY SECTION IN CENTER IS REACTOR BUILDING. TWO-STORY CONTROL ROOM AND OFFICE BUILDING, TRA-647, IS BETWEEN IT AND MTR SERVICE BUILDING. STRUCTURE TO THE LEFT (WITH NO FRAMING YET) IS COMPRESSOR BUILDING, TRA-643, AND BEYOND IT WILL BE HEAT EXCHANGER BUILDING, TRA-644, GREAT SOUTHERN BUTTE ON HORIZON. INL NEGATIVE NO. 56-2382. Jack L. Anderson, Photographer, 6/10/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

    SciTech Connect

    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.

  9. Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor

    SciTech Connect

    Bess, John D.; Fujimoto, Nozomu

    2014-10-09

    Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the HTTR. Additional measurements of a fully-loaded subcritical configuration, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely-detailed models of the HTTR as much of the design information is still proprietary. However, the uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that comprise the HTTR. Monte Carlo calculations of keff are between approximately 0.9 % and 2.7 % greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulation of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  10. Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor

    DOE PAGES

    Bess, John D.; Fujimoto, Nozomu

    2014-10-09

    Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the HTTR. Additional measurements of a fully-loaded subcritical configuration, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely-detailed models of the HTTR as much of the design information is still proprietary. However, the uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. Dominant uncertainties in themore » experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that comprise the HTTR. Monte Carlo calculations of keff are between approximately 0.9 % and 2.7 % greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulation of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

  11. High Flux Isotopes Reactor (HFIR) Cooling Towers Demolition Waste Management

    SciTech Connect

    Pudelek, R. E.; Gilbert, W. C.

    2002-02-26

    This paper describes the results of a joint initiative between Oak Ridge National Laboratory, operated by UT-Battelle, and Bechtel Jacobs Company, LLC (BJC) to characterize, package, transport, treat, and dispose of demolition waste from the High Flux Isotope Reactor (HFIR), Cooling Tower. The demolition and removal of waste from the site was the first critical step in the planned HFIR beryllium reflector replacement outage scheduled. The outage was scheduled to last a maximum of six months. Demolition and removal of the waste was critical because a new tower was to be constructed over the old concrete water basin. A detailed sampling and analysis plan was developed to characterize the hazardous and radiological constituents of the components of the Cooling Tower. Analyses were performed for Resource Conservation and Recovery Act (RCRA) heavy metals and semi-volatile constituents as defined by 40 CFR 261 and radiological parameters including gross alpha, gross beta, gross gamma, alpha-emitting isotopes and beta-emitting isotopes. Analysis of metals and semi-volatile constituents indicated no exceedances of regulatory limits. Analysis of radionuclides identified uranium and thorium and associated daughters. In addition 60Co, 99Tc, 226Rm, and 228Rm were identified. Most of the tower materials were determined to be low level radioactive waste. A small quantity was determined not to be radioactive, or could be decontaminated. The tower was dismantled October 2000 to January 2001 using a detailed step-by-step process to aid waste segregation and container loading. The volume of waste as packaged for treatment was approximately 1982 cubic meters (70,000 cubic feet). This volume was comprised of plastic ({approx}47%), wood ({approx}38%) and asbestos transite ({approx}14%). The remaining {approx}1% consisted of the fire protection piping (contaminated with lead-based paint) and incidental metal from conduit, nails and braces/supports, and sludge from the basin. The waste

  12. High-purity isotopes production by short-term HFIR (High-Flux Isotope Reactor) irradiations

    SciTech Connect

    Knauer, J.B.; Alexander, C.W.; Bigelow, J.E.; Wiggins, J.T.

    1987-01-01

    The Transuranium Processing Plant (TPP) at Oak Ridge National Laboratory, along with the High-Flux Isotope Reactor (HFIR) were built to produce the quantities of the transplutonium elements needed for the heavy-element research programs of the US Department of Energy (USDOE). This document consists of viewographs and tables.

  13. Evaluation of HFIR (High Flux Isotope Reactor) pressure-vessel integrity considering radiation embrittlement

    SciTech Connect

    Cheverton, R.D.; Merkle, J.G.; Nanstad, R.K.

    1988-04-01

    The High Flux Isotope Reactor (HFIR) pressure vessel has been in service for 20 years, and during this time, radiation damage was monitored with a vessel-material surveillance program. In mid-November 1986, data from this program indicated that the radiation-induced reduction in fracture toughness was greater than expected. As a result, a reevaluation of vessel integrity was undertaken. Updated methods of fracture-mechanics analysis were applied, and an accelerated irradiations program was conducted using the Oak Ridge Research Reactor. Results of these efforts indicate that (1) the vessel life can be extended 10 years if the reactor power level is reduced 15% and if the vessel is subjected to a hydrostatic proof test each year; (2) during the 10-year life extension, significant radiation damage will be limited to a rather small area around the beam tubes; and (3) the greater-than-expected damage rate is the result of the very low neutron flux in the HFIR vessel relative to that in samples of material irradiated in materials-testing reactors (a factor of approx.10/sup 4/ less), that is, a rate effect.

  14. Design Study for a Low-enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2007

    SciTech Connect

    Primm, Trent; Ellis, Ronald James; Gehin, Jess C; Ilas, Germina; Miller, James Henry; Sease, John D

    2007-11-01

    This report documents progress made during fiscal year 2007 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low enriched uranium fuel (LEU). Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. A high volume fraction U/Mo-in-Al fuel could attain the same neutron flux performance as with the current, HEU fuel but materials considerations appear to preclude production and irradiation of such a fuel. A diffusion barrier would be required if Al is to be retained as the interstitial medium and the additional volume required for this barrier would degrade performance. Attaining the high volume fraction (55 wt. %) of U/Mo assumed in the computational study while maintaining the current fuel plate acceptance level at the fuel manufacturer is unlikely, i.e. no increase in the percentage of plates rejected for non-compliance with the fuel specification. Substitution of a zirconium alloy for Al would significantly increase the weight of the fuel element, the cost of the fuel element, and introduce an as-yet untried manufacturing process. A monolithic U-10Mo foil is the choice of LEU fuel for HFIR. Preliminary calculations indicate that with a modest increase in reactor power, the flux performance of the reactor can be maintained at the current level. A linearly-graded, radial fuel thickness profile is preferred to the arched profile currently used in HEU fuel because the LEU fuel media is a metal alloy foil rather than a powder. Developments in analysis capability and nuclear data processing techniques are underway with the goal of verifying the preliminary calculations of LEU flux performance. A conceptual study of the operational cost of an LEU fuel fabrication facility yielded the conclusion that the annual fuel cost to the HFIR would increase significantly from the current, HEU fuel cycle. Though manufacturing can be accomplished with existing technology

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

    NASA Astrophysics Data System (ADS)

    Fourmentel, D.; Filliatre, P.; Villard, J. F.; Lyoussi, A.; Reynard-Carette, C.; Carcreff, H.

    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-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.

  16. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    SciTech Connect

    Not Available

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report.

  17. Irradiation experiments on high temperature gas-cooled reactor fuels and graphites at the high flux reactor petten

    NASA Astrophysics Data System (ADS)

    Ahlf, J.; Conrad, R.; Cundy, M.; Scheurer, H.

    1990-04-01

    Because of its favourable design and operational characteristics and the availability of dedicated experimental equipment the High Flux Reactor at Petten has been extensively used as a test bed for HTR fuel and graphite irradiations for more than 20 years. Earlier fuel testing programmes contributed to the development of the coated fuel particle concept by extended screening tests. Now these programmes concentrate on performance testing of reference coated fuel particles and reference fuel elements for the German HTR-Module, the HTR-500 and to a lesser extent for the US HTGR concepts. It is shown with representative examples that these fuels have excellent fission product retention capabilities under normal and anticipated off-normal operating conditions. Extended irradiation programmes in the HFR Petten have significantly contributed to the database for the design of HTR graphite structures. The programmes not only comprise radiation damage accumulation in the temperature range from 570 to 1570 K up to very high fast neutron fluences and its influence on technological properties, but also irradiations under specified load conditions to investigate the irradiation creep behaviour of various graphites in the temperature range 570 to 1170 K.

  18. Dissolution Flowsheet for High Flux Isotope Reactor Fuel

    SciTech Connect

    Daniel, W. E.; Rudisill, T. S.; O'Rourke, P. E.; Karay, N. S

    2016-09-27

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy, and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H2 and other permanent gases in the dissolution offgas, allowing the development of H2 generation rate profiles. The H2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

  19. ENGINEERING TEST REACTOR (ETR) BUILDING, TRA642. CONTEXTUAL VIEW, CAMERA FACING ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    ENGINEERING TEST REACTOR (ETR) BUILDING, TRA-642. CONTEXTUAL VIEW, CAMERA FACING EAST. VERTICAL METAL SIDING. ROOF IS SLIGHTLY ELEVATED AT CENTER LINE FOR DRAINAGE. WEST SIDE OF ETR COMPRESSOR BUILDING, TRA-643, PROJECTS TOWARD LEFT AT FAR END OF ETR BUILDING. INL NEGATIVE NO. HD46-37-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  20. Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor

    SciTech Connect

    Ilas, Germina; Primm, Trent

    2011-05-01

    An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

  1. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    SciTech Connect

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  2. Apparatus for high flux photocatalytic pollution control using a rotating fluidized bed reactor

    DOEpatents

    Tabatabaie-Raissi, Ali; Muradov, Nazim Z.; Martin, Eric

    2003-06-24

    An apparatus based on optimizing photoprocess energetics by decoupling of the process energy efficiency from the DRE for target contaminants. The technique is applicable to both low- and high-flux photoreactor design and scale-up. An apparatus for high-flux photocatalytic pollution control is based on the implementation of multifunctional metal oxide aerogels and other media in conjunction with a novel rotating fluidized particle bed reactor.

  3. The High-Resolution Powder Diffractometer at the High Flux Isotope Reactor

    SciTech Connect

    Garlea, Vasile O; Chakoumakos, Bryan C; Moore, Scott A; Taylor, Gerald Brent; Chae, Timothy L; Maples, Ron G; Riedel, Richard A; Lynn, Gary W; Selby, Douglas L

    2010-01-01

    Neutron powder diffraction is increasingly recognized as one of the most powerful techniques for studying the structural and magnetic properties of advanced materials. Despite the growing demand to study an ever-increasing array of interesting materials, there is only a handful of neutron diffractometers available to serve the U.S. neutron scattering community. This article describes the new high-resolution powder diffractometer that has recently been installed at the High Flux Isotope Reactor in Oak Ridge. The instrument is designed to provide an optimum balance between high neutron flux and high resolution. Due to its versatility the diffractometer can be employed for a large variety of experiments, but it is particularly adapted for refinements of structures with large interplanar spacings as well as of complex magnetic structures. In addition to traditional crystal and magnetic structural refinements, studies of phase transitions, thermal expansion, texture analysis, and ab initio structure solution from powder data can be undertaken.

  4. The Decontamination, Decommissioning, and Demolition of the Engineering Test Reactor at the Idaho Cleanup Project

    SciTech Connect

    Coyne, D.W.

    2008-07-01

    In September 2007, CH2M-WG Idaho completed the decontamination, decommissioning and demolition (D and D) of the Engineering Test Reactor (ETR) facility. The 50-year-old research reactor, located at the Idaho National Laboratory site, posed significant challenges involving regulations governing the demolition of a historical facility, the removal of a large amount of hazardous materials as well as issues associated with the removal and disposal of the 112-ton reactor vessel. Prior to commencing full scale D and D, hazardous constituents were removed including cadmium, PCB oils and electrical components, lead, asbestos and mercury among others. The reactor required isolation in order to be removed. Due to activated metal within the reactor vessel, dose rates in the core region were approximately 1100 R/hr. Subsequent dose rates outside the vessel varied from 60 mR to greater than 2 R. Due to the dose rates, the project team decided to fill the reactor vessel with grout to a level above the core region and below the discharge to the canal. To remove the reactor, access to the 17 mounting shoes was required. These shoes were encased in the high density concrete biological shield approximately 8 feet below grade. The project team used explosives to remove the biological shield. The demolition had to be controlled to prevent damaging the reactor vessel and to limit the seismic impact on a nearby operating reactor. Upon completion of the blast, the concrete was removed exposing the support shoes for the vessel. The reactor building was then demolished to accommodate the twin gantry system used to lift the reactor vessel. In September, the reactor vessel was lifted and placed onto a multi-axle trailer for transport to an onsite disposal facility. (authors)

  5. Production of transplutonium elements in the high flux isotope reactor

    SciTech Connect

    Bigelow, J.E.; Corbett, B.L.; King, L.J.; McGuire, S.C.; Sims, T.M.

    1981-01-01

    The techniques described here have been demonstrated to predict the contents of transplutonium element production targets, at least for isotopes of mass 253 or less. The HFIR irradiation model is a workhorse for planning the TRU processing campaigns, for certifying the heat evolution rate of targets prior to insertion in the reactor, for predicting future production capabilities over a multi-year period, and for making optimization studies. Practical considerations, however, may limit the range of available options so that optimum operation is not always achievable. We do intend, however, to keep fine-tuning the constants which define the cross sections as time permits. We need to do more work on optimizing the production of /sup 250/Cm, /sup 254/Es, /sup 255/Es, and ultimately /sup 257/Fm, since researchers are interested in obtaining larger quantities of these rare and difficult-to-produce nuclides. 7 figures, 2 tables.

  6. FASTER Test Reactor Preconceptual Design Report

    SciTech Connect

    Grandy, C.; Belch, H.; Brunett, A. J.; Heidet, F.; Hill, R.; Hoffman, E.; Jin, E.; Mohamed, W.; Moisseytsev, A.; Passerini, S.; Sienicki, J.; Sumner, T.; Vilim, R.; Hayes, S.

    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.

  7. Probability of fracture and life extension estimate of the high-flux isotope reactor vessel

    SciTech Connect

    Chang, S.J.

    1998-08-01

    The state of the vessel steel embrittlement as a result of neutron irradiation can be measured by its increase in ductile-brittle transition temperature (DBTT) for fracture, often denoted by RT{sub NDT} for carbon steel. This transition temperature can be calibrated by the drop-weight test and, sometimes, by the Charpy impact test. The life extension for the high-flux isotope reactor (HFIR) vessel is calculated by using the method of fracture mechanics that is incorporated with the effect of the DBTT change. The failure probability of the HFIR vessel is limited as the life of the vessel by the reactor core melt probability of 10{sup {minus}4}. The operating safety of the reactor is ensured by periodic hydrostatic pressure test (hydrotest). The hydrotest is performed in order to determine a safe vessel static pressure. The fracture probability as a result of the hydrostatic pressure test is calculated and is used to determine the life of the vessel. Failure to perform hydrotest imposes the limit on the life of the vessel. The conventional method of fracture probability calculations such as that used by the NRC-sponsored PRAISE CODE and the FAVOR CODE developed in this Laboratory are based on the Monte Carlo simulation. Heavy computations are required. An alternative method of fracture probability calculation by direct probability integration is developed in this paper. The present approach offers simple and expedient ways to obtain numerical results without losing any generality. In this paper, numerical results on (1) the probability of vessel fracture, (2) the hydrotest time interval, and (3) the hydrotest pressure as a result of the DBTT increase are obtained.

  8. Cold source moderator vessel development for the High Flux Isotope Reactor: Thermal-hydraulic studies

    SciTech Connect

    Williams, P.T.; Lucas, A.T.; Wendel, M.W.

    1998-07-01

    A project is underway at Oak Ridge National Laboratory (ORNL) to design, test, and install a cold neutron source facility in the High Flux Isotope Reactor (HFIR). This new cold source employs supercritical hydrogen at cryogenic temperatures both as the medium for neutron moderation and as the working fluid for removal of internally-generated nuclear heating. The competing design goals of minimizing moderator vessel mass and providing adequate structural integrity for the vessel motivated the requirement of detailed multidimensional thermal-hydraulic analyses of the moderator vessel as a critical design subtask. This paper provides a summary review of the HFIR cold source moderator vessel design and a description of the thermal-hydraulic studies that were carried out to support the vessel development.

  9. Decommissioning of the High Flux Beam Reactor at Brookhaven National Laboratory.

    PubMed

    Hu, Jih-Perng; Reciniello, Richard N; Holden, Norman E

    2012-08-01

    The High Flux Beam Reactor (HFBR) at the Brookhaven National Laboratory was a heavy-water cooled and moderated reactor that achieved criticality on 31 October 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost 3 y for safety and environmental reviews. In November 1999, the United States Department of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR's cleanup performed during 1999-2009, to allow the BNL facilities to be re-accessed by the public, will be described in the paper.

  10. STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS

    SciTech Connect

    Katoh, Yutai; Koyanagi, Takaaki; Kiggans, Jim; Cetiner, Nesrin; McDuffee, Joel

    2014-09-01

    Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

  11. Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

    DOE PAGES

    Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.; ...

    2016-12-21

    Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate

  12. Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

    SciTech Connect

    Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.; Carbajo, Juan J.; Greenwood, M. Scott; Hale, Richard; Harrison, Thomas J.; Powers, Jeffrey J.; Robb, Kevin R.; Terrell, Jerry; Wysocki, Aaron J.; Gehin, Jess C.; Worrall, Andrew

    2016-12-21

    Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It

  13. Establishing a Cost Basis for Converting the High Flux Isotope Reactor from High Enriched to Low Enriched Uranium Fuel

    SciTech Connect

    Primm, Trent; Guida, Tracey

    2010-02-01

    Under the auspices of the Global Threat Reduction Initiative Reduced Enrichment for Research and Test Reactors Program, the National Nuclear Security Administration /Department of Energy (NNSA/DOE) has, as a goal, to convert research reactors worldwide from weapons grade to non-weapons grade uranium. The High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL) is one of the candidates for conversion of fuel from high enriched uranium (HEU) to low enriched uranium (LEU). A well documented business model, including tasks, costs, and schedules was developed to plan the conversion of HFIR. Using Microsoft Project, a detailed outline of the conversion program was established and consists of LEU fuel design activities, a fresh fuel shipping cask, improvements to the HFIR reactor building, and spent fuel operations. Current-value costs total $76 million dollars, include over 100 subtasks, and will take over 10 years to complete. The model and schedule follows the path of the fuel from receipt from fuel fabricator to delivery to spent fuel storage and illustrates the duration, start, and completion dates of each subtask to be completed. Assumptions that form the basis of the cost estimate have significant impact on cost and schedule.

  14. High heat flux issues for plasma-facing components in fusion reactors

    NASA Astrophysics Data System (ADS)

    Watson, Robert D.

    1993-02-01

    Plasma facing components in tokamak fusion reactors are faced with a number of difficult high heat flux issues. These components include: first wall armor tiles, pumped limiters, diverter plates, rf antennae structure, and diagnostic probes. Peak heat fluxes are 15 - 30 MW/m2 for diverter plates, which will operate for 100 - 1000 seconds in future tokamaks. Disruption heat fluxes can approach 100,000 MW/m2 for 0.1 ms. Diverter plates are water-cooled heat sinks with armor tiles brazed on to the plasma facing side. Heat sink materials include OFHC, GlidcopTM, TZM, Mo-41Re, and niobium alloys. Armor tile materials include: carbon fiber composites, beryllium, silicon carbide, tungsten, and molybdenum. Tile thickness range from 2 - 10 mm, and heat sinks are 1 - 3 mm. A twisted tape insert is used to enhance heat transfer and increase the burnout safety margin from critical heat flux limits to 50 - 60 MW/m2 with water at 10 m/s and 4 MPa. Tests using rastered electron beams have shown thermal fatigue failures from cracks at the brazed interface between tiles and the heat sink after only 1000 cycles at 10 - 15 MW/m2. These fatigue lifetimes need to be increased an order of magnitude to meet future requirements. Other critical issues for plasma facing components include: surface erosion from sputtering and disruption erosion, eddy current forces and runaway electron impact from disruptions, neutron damage, tritium retention and release, remote maintenance of radioactive components, corrosion-erosion, and loss-of-coolant accidents.

  15. FASTER test reactor preconceptual design report summary

    SciTech Connect

    Grandy, C.; Belch, H.; Brunett, A.; Heidet, F.; Hill, R.; Hoffman, E.; Jin, E.; Mohamed, W.; Moisseytsev, A.; Passerini, S.; Sienicki, J.; Sumner, T.; Vilim, R.; Hayes, Steven

    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.

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

    SciTech Connect

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

    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.

  17. High heat flux engineering in solar energy applications

    SciTech Connect

    Cameron, C.P.

    1993-07-01

    Solar thermal energy systems can produce heat fluxes in excess of 10,000 kW/m{sup 2}. This paper provides an introduction to the solar concentrators that produce high heat flux, the receivers that convert the flux into usable thermal energy, and the instrumentation systems used to measure flux in the solar environment. References are incorporated to direct the reader to detailed technical information.

  18. A conceptual high flux reactor design with scope for use in ADS applications

    NASA Astrophysics Data System (ADS)

    Pal, Usha; Jagannathan, V.

    2007-02-01

    A 100 MWt reactor design has been conceived to support flux level of the order of 1015 n/cm2/s in selected flux trap zones. The physics design considers high enriched metallic alloy fuel in the form of annular plates placed in a D2O moderator tank in a hexagonal lattice arrangement. By choosing a tight lattice pitch in the central region and double the lattice pitch in the outer region, it is possible to have both high fast flux and thermal flux trap zones. By design the flux level in the seed fuel has been kept lower than in the high flux trap zones so that the burning rate of the seed is reduced. Another important objective of the design is to maximize the time interval of refueling. As against a typical refueling interval of a few weeks in such high flux reactor cores, it is desired to maximize this period to as much as six months or even one year. This is possible to achieve by eliminating the conventional control absorbers and replacing them with a suitable amount of fertile material loading in the reactor. Requisite number of seedless thorium-aluminum alloy plates are placed at regular lattice locations vacated by seed fuel in alternate fuel layers. It is seen that these thorium plates are capable of acquiring asymptotic fissile content of 14 g/kg in about 100 days of irradiation at a flux level of 8 x 1014 n/cm2/s. In summary, the core has a relatively higher fast flux in the central region and high thermal flux in the outer region. The present physics design envisages a flat core excess reactivity for the longest possible cycle length of 6 months to one year. It is also possible to modify the design for constant subcriticality for about the same period or longer duration by considering neutron spallation source at the centre and curtailing the power density in the inner core region by shielding it with a layer of thoria fuel loading.

  19. Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor

    SciTech Connect

    Primm, Trent; Gehin, Jess C

    2009-04-01

    A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

  20. Neutron Spectral Brightness of Cold Guide 4 at the High Flux Isotope Reactor

    NASA Astrophysics Data System (ADS)

    Winn, B. L.; Robertson, J. L.; Iverson, E. B.; Selby, D. L.

    2010-11-01

    The High Flux Isotope Reactor resumed operation in June of 2007 with a supercritical hydrogen cold source in horizontal beam tube 4. Cold guide 4 is a guide system designed to deliver neutrons from this source with a reasonable flux at wavelengths greater than 4 Å to several instruments, and includes a 15-m, 96-section, 4-channel bender. A time-of-flight spectrum with calibrated detector was recorded at port C of cold guide 4, and compared to McStas simulations, to generate a brightness spectrum.

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

    SciTech Connect

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

    2012-01-01

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

  2. Studies of Plutonium-238 Production at the High Flux Isotope Reactor

    SciTech Connect

    Lastres, Oscar; Chandler, David; Jarrell, Joshua J; Maldonado, G. Ivan

    2011-01-01

    The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is a versatile 85 MW{sub th}, pressurized, light water-cooled and -moderated research reactor. The core consists of two fuel elements, an inner fuel element (IFE) and an outer fuel element (OFE), each constructed of involute fuel plates containing high-enriched-uranium (HEU) fuel ({approx}93 wt% {sup 235}U/U) in the form of U{sub 3}O{sub 8} in an Al matrix and encapsulated in Al-6061 clad. An over-moderated flux trap is located in the center of the core, a large beryllium reflector is located on the outside of the core, and two control elements (CE) are located between the fuel and the reflector. The flux trap and reflector house numerous experimental facilities which are used for isotope production, material irradiation, and cold/thermal neutron scattering. Over the past five decades, the US Department of Energy (DOE) and its agencies have been producing radioisotope power systems used by the National Aeronautics and Space Administration (NASA) for unmanned, long-term space exploration missions. Plutonium-238 is used to power Radioisotope Thermoelectric Generators (RTG) because it has a very long half-life (t{sub 1/2} {approx} 89 yr.) and it generates about 0.5 watts/gram when it decays via alpha emission. Due to the recent shortage and uncertainty of future production, the DOE has proposed a plan to the US Congress to produce {sup 238}Pu by irradiating {sup 237}Np as early as in fiscal year 2011. An annual production rate of 1.5 to 2.0 kg of {sup 238}Pu is expected to satisfy these needs and could be produced in existing national nuclear facilities like HFIR and the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Reactors at the Savannah River Site were used in the past for {sup 238}Pu production but were shut down after the last production in 1988. The nation's {sup 237}Np inventory is currently stored at INL. A plan for producing {sup 238}Pu at US research reactor

  3. Development of advanced high-temperature heat flux sensors. Phase 2: Verification testing

    NASA Technical Reports Server (NTRS)

    Atkinson, W. H.; Cyr, M. A.; Strange, R. R.

    1985-01-01

    A two-phase program is conducted to develop heat flux sensors capable of making heat flux measurements throughout the hot section of gas turbine engines. In Phase 1, three types of heat flux sensors are selected; embedded thermocouple, laminated, and Gardon gauge sensors. A demonstration of the ability of these sensors to operate in an actual engine environment is reported. A segmented liner of each of two combustors being used in the Broad Specification Fuels Combustor program is instrumented with the three types of heat flux sensors then tested in a high pressure combustor rig. Radiometer probes are also used to measure the radiant heat loads to more fully characterize the combustor environment. Test results show the heat flux sensors to be in good agreement with radiometer probes and the predicted data trends. In general, heat flux sensors have strong potential for use in combustor development programs.

  4. The HB-2D Polarized Neutron Development Beamline at the High Flux Isotope Reactor

    NASA Astrophysics Data System (ADS)

    Crow, Lowell; Hamilton, WA; Zhao, JK; Robertson, JL

    2016-09-01

    The Polarized Neutron Development beamline, recently commissioned at the HB-2D position on the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory, provides a tool for development and testing of polarizers, polarized neutron devices, and prototyping of polarized neutron techniques. With available monochromators including pyrolytic graphite and polarizing enriched Fe-57 (Si), the instrument has operated at 4.25 and 2.6 Å wavelengths, using crystal, supermirror, or He-3 polarizers and analyzers in various configurations. The Neutron Optics and Development Team has used the beamline for testing of He-3 polarizers for use at other HFIR and Spallation Neutron Source (SNS) instruments, as well as a variety of flipper devices. Recently, we have acquired new supermirror polarizers which have improved the instrument performance. The team and collaborators also have continuing demonstration experiments of spin-echo focusing techniques, and plans to conduct polarized diffraction measurements. The beamline is also used to support a growing use of polarization techniques at present and future instruments at SNS and HFIR.

  5. RADIATION DOSIMETRY OF THE PRESSURE VESSEL INTERNALS OF THE HIGH FLUX BEAM REACTOR.

    SciTech Connect

    HOLDEN,N.E.; RECINIELLO,R.N.; HU,J.P.; RORER,D.C.

    2002-08-18

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the transition plate, and the control rod blades. The measurements were made using Red Perspex{trademark} polymethyl methacrylate high-level film dosimeters, a Radcal ''peanut'' ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rate, the Monte Carlo MCNP code and geometric progressive Microshield code were used to model the gamma transport and dose buildup.

  6. The Role of COMSOL Toward a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor

    SciTech Connect

    Freels, James D; Arimilli, Rao V; Lowe, Kirk T; Bodey, Isaac T

    2009-01-01

    Design and safety analyses are underway to convert the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) from a high-enriched uranium (HEU) fuel to a low-enriched uranium (LEU) fuel. The primary constraint for the project is that the overall fuel plate dimensions and the current neutron flux performance must remain unchanged. This allows minimal impact on the facility and cost for the conversion, and provides transparency to the HFIR customer base and research projects that depend on the facility for isotopes and neutron flux. As a consequence, the LEU design demands more accuracy and less margin in the analysis efforts than the original design. Several technical disciplines are required to complete this conversion including nuclear reactor physics, heat transfer, fluid dynamics, structural mechanics, fuel fabrication, and engineering design. The role of COMSOL is to provide the fully-coupled 3D multi-physics analysis for heat transfer, turbulent flow, and structural mechanics of the fuel plates and flow channels. A goal is for COMSOL to simulate the entire fuel element array of fuel plates (171 inner, 369 outer). This paper describes the progress that has been made toward development of benchmark validation models of the existing HEU inner-element fuel plates.

  7. Establishing Specifications for Low Enriched Uranium Fuel Operations Conducted Outside the High Flux Isotope Reactor Site

    SciTech Connect

    Pinkston, Daniel; Primm, Trent; Renfro, David G; Sease, John D

    2010-10-01

    The National Nuclear Security Administration (NNSA) has funded staff at Oak Ridge National Laboratory (ORNL) to study the conversion of the High Flux Isotope Reactor (HFIR) from the current, high enriched uranium fuel to low enriched uranium fuel. The LEU fuel form is a metal alloy that has never been used in HFIR or any HFIR-like reactor. This report provides documentation of a process for the creation of a fuel specification that will meet all applicable regulations and guidelines to which UT-Battelle, LLC (UTB) the operating contractor for ORNL - must adhere. This process will allow UTB to purchase LEU fuel for HFIR and be assured of the quality of the fuel being procured.

  8. Reactor physics input to the safety analysis report for the High Flux Isotope Reactor

    SciTech Connect

    Primm, R.T. III.

    1992-03-01

    HFIR specific, few group neutron and coupled neutron-gamma libraries have been prepared. These are based on data from ENDF/B-V and beginning-of-life (BOL) conditions. The neutron library includes actinide data for curium target rods. Six critical experiments, collectively designated HFIR critical experiment 4, were analyzed. Calculated k-effective was 2% high at BOL-typical conditions but was 1.0 at end-of-life-typical conditions. The local power density distributions were calculated for each of the critical experiments. The axially averaged values at a given radius were frequently within experimental error. However at individual points, the calculated local power densities were significantly different from the experimentally derived values (several times greater than experimental uncertainty). A reassessment of the foil activation data with transport theory techniques seems desirable. Using the results of the critical experiments study, a model of current HFIR configuration was prepared. As with the critical experiments, BOL k-effective was high (3%). However, end-of-life k-effective was high (2%). The end-of-life concentrations of fission products were compared to those generated using the ORIGEN code. Agreement was generally good through differences in the inventories of some important nuclides, Xe and I, need to be understood. End-of-cycle curium target isotopics based on measured, discharged target rods were compared to calculated values and agreement was good. Axial flux plots at various irradiation positions were generated. Time-dependent power distributions based on two-dimensional calculations were provided.

  9. High heat flux testing capabilities at Sandia National Laboratories - New Mexico

    SciTech Connect

    Youchison, D.L.; McDonald, J.M.; Wold, L.S.

    1994-12-31

    High heat flux testing for the United States fusion power program is the primary mission of the Plasma Materials Test Facility (PMTF) located at Sandia National Laboratories - New Mexico. This facility, which is owned by the United States Department of Energy, has been in operation for over 17 years and has provided much of the high heat flux data used in the design and evaluation of plasma facing components for many of the world`s magnetic fusion, tokamak experiments. In addition to domestic tokamaks such as Tokamak Fusion Test Reactor (TFTR) at Princeton and the DIII-D tokamak at General Atomics, components for international experiments like TEXTOR, Tore-Supra, and JET also have been tested at the PMTF. High heat flux testing spans a wide spectrum including thermal shock tests on passively cooled materials, thermal response and thermal fatigue tests on actively cooled components, critical heat flux-burnout tests, braze reliability tests and safety related tests. The objective of this article is to provide a brief overview of the high heat flux testing capabilities at the PMTF and describe a few of the experiments performed over the last year.

  10. A neutronic feasibility study for LEU conversion of the high flux isotope reactor (HFIR).

    SciTech Connect

    Mo, S. C.

    1998-01-14

    A neutronic feasibility study was performed to determine the uranium densities that would be required to convert the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) from HEU (93%) to LEU (<20%)fuel. The LEU core that was studied is the same as the current HEU core, except for potential changes in the design of the fuel plates. The study concludes that conversion of HFIR from HEU to LEU fuel would require an advanced fuel with a uranium density of 6-7 gU/cm{sup 3} in the inner fuel element and 9-10 gU/cm{sup 3} in the outer fuel element to match the cycle length of the HEU core. LEU fuel with uranium density up to 4.8 gU/cm{sup 3} is currently qualified for research reactor use. Modifications in fuel grading and burnable poison distribution are needed to produce an acceptable power distribution.

  11. Two dimensional impinging jet cooling of high heat flux surface in magnetic confinement fusion reactor

    SciTech Connect

    Inoue, A.; Tanno, T.; Takahashi, M.

    1994-12-31

    Divertor surface of a magnetic confinement fusion reactor is exposed to strong radiation heating by high flux charged particles. According to standard design of the ITER, the heat flux of the divertor surface becomes average 15MW/m{sup 2} or more. In this study, a cooling by a two dimensional impinging jet flow is proposed to cool such high heat flux surface. For an impinging jet flow to flat heated surface, such as CHF is obtained only in the limited surface region where the jet flow hits directly. Apart from the region, the CHF decreases abruptly with the distance from the center. The main reason is that the pressure decreases abruptly as apart from the center region and the liquid flow is spread away from the heated surface region by the strong boiling. To overcome these difficulties, the authors propose that the impinging jet is applied to the heat transfer wall with a concave surface, because the pressure change becomes mild by the centrifugal force along the curved surface. The increase of the radial pressure gradient in the vertical direction to the curved surface promotes the departure of vapor bubbles near the wall region. It is expected that this mechanism as well as keeping high pressure along the flow works to enhance the CHF. To obtain the high heat flux in the wide region, a use of a two-dimensional impinging jet is suitable instead of a round jet.

  12. Response of materials to high heat fluxes during operation in fusion reactors

    SciTech Connect

    Hassanein, A.M.

    1988-07-01

    Very high energy deposition on first wall and other components of a fusion reactor is expected due to plasma instabilities during both normal and off-normal operating conditions. Off-normal operating conditions result from plasma disruptions where the plasma loses confinement and dumps its energy on the reactor components. High heat flux may also result from normal operating conditions due to fluctuations in plasma edge conditions. This high energy dump in a short time results in very high surface temperatures and may consequently cause melting and vaporization of these materials. The net erosion rates resulting from melting and vaporization are very important to estimate the lifetime of such components. The response of different candidate materials to this high heat fluxes is determined for different energy densities and deposition times. The analysis used a previously developed model to solve the heat conduction equation in two moving boundaries. One moving boundary is at the surface to account for surface recession due to vaporization and the second moving boundary is to account for the solid-liquid interface inside the material. The calculations are done parametrically for both the expected energy deposited and the deposition time. These ranges of energy and time are based on recent experimental observations in current fusion devices. The candidate materials analyzed are stainless steel, carbon, and tungsten. 8 refs., 9 figs.

  13. Evaluation of some candidate materials for automobile thermal reactors in engine-dynamometer screening tests

    NASA Technical Reports Server (NTRS)

    Oldrieve, R. E.

    1971-01-01

    Fourteen materials were evaluated in engine screening tests on full-size thermal reactors for automobile engine pollution control systems. Cyclic test-stand engine operation provided 2 hours at 1040 C and a 20-minute air-cool to 70 C each test cycle. Each reactor material was exposed to 83 cycles in 200 hours of engine testing. On the basis of resistance to oxidation and distortion, the best materials included two ferritic iron alloys (Ge 1541 and Armco 18S/R), several commercial oxidation-resistant coatings on AlSl 651 (19-9 DL), and possibly uncoated AISI 310. The best commercial coatings were Cr-Al, Ni-Cr, and a glass ceramic.

  14. Development of a Scale Model for High Flux Isotope Reactor Cycle 400

    SciTech Connect

    Ilas, Dan

    2012-03-01

    The development of a comprehensive SCALE computational model for the High Flux Isotope Reactor (HFIR) is documented and discussed in this report. The SCALE model has equivalent features and functionality as the reference MCNP model for Cycle 400 that has been used extensively for HFIR safety analyses and for HFIR experiment design and analyses. Numerical comparisons of the SCALE and MCNP models for the multiplication constant, power density distribution in the fuel, and neutron fluxes at several locations in HFIR indicate excellent agreement between the results predicted with the two models. The SCALE HFIR model is presented in sufficient detail to provide the users of the model with a tool that can be easily customized for various safety analysis or experiment design requirements.

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

    SciTech Connect

    Sterbentz, James William; Bayless, Paul David

    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.

  16. DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

    SciTech Connect

    Cook, David Howard; Freels, James D; Ilas, Germina; Jolly, Brian C; Miller, James Henry; Primm, Trent; Renfro, David G; Sease, John D; Pinkston, Daniel

    2011-02-01

    This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

  17. Second annual progress report on United States-Japan collaborative testing in the High Flux Isotope Reactor and the Oak Ridge Research Reactor for the period ending September 30, 1985

    SciTech Connect

    Scott, J.L.; Grossbeck, M.L.; Mansur, L.K.; Rowcliffe, A.F.; Siman-Tov, I.I.; Thoms, K.R.; Tanaka, M.P.; Hamada, S.; Kendo, T.; Hishinuma, A.

    1986-08-01

    The second year of the program of US-Japan collaborative testing in the HFIR and ORR has been successfully completed. Two of eight phase-I target capsules were irradiated, and postirradiation testing was begun. Two spectral-tailoring capsules, MFE-6J and -7J, were fabricated and installed in the ORR. The JEOL JEM 2000FX microscope was installed at ORNL and is now being operated routinely. Microstructural data of the JPCA in the SA, 10%, and 20% cold-worked conditions and type J316 in the SA and 20% cold-worked conditions reveal that all specimens examined clearly show a high concentration of fine helium bubbles after irradiation to about 30 dpa at 300/sup 0/C (in HFIR). Precipitation of MC was observed in 20% cold-worked JPCA. Swelling of all specimens was less than 1%.

  18. Status of FeCrAl ODS Irradiations in the High Flux Isotope Reactor

    SciTech Connect

    Field, Kevin G.; Howard, Richard H.

    2016-08-19

    FeCrAl oxide-dispersion strengthened (ODS) alloys are an attractive sub-set alloy class of the more global FeCrAl material class for nuclear applications due to their high-temperature steam oxidation resistance and hypothesized enhanced radiation tolerance. A need currently exists to determine the radiation tolerance of these newly developed alloys. To address this need, a preliminary study was conducted using the High Flux Isotope Reactor (HFIR) to irradiate an early generation FeCrAl ODS alloy, 125YF. Preliminary post-irradiation examination (PIE) on these irradiated specimens have shown good radiation tolerance at elevated temperatures (≥330°C) but possible radiation-induced hardening and embrittlement at irradiations of 200°C to a damage level of 1.9 displacement per atom (dpa). Building on this experience, a new series of irradiations are currently being conceptualized. This irradiation series called the FCAD irradiation program will irradiate the latest generation FeCrAl ODS and FeCr ODS alloys to significantly higher doses. These experiments will provide the necessary information to determine the mechanical performance of irradiated FeCrAl ODS alloys at light water reactor and fast reactor conditions.

  19. Potential role of the Fast Flux Test Facility and the advanced test reactor in the U.S. tritium production system

    SciTech Connect

    Dautel, W.A.

    1996-10-01

    The Deparunent of Energy is currently engaged in a dual-track strategy to develop an accelerator and a conunercial light water reactor (CLWR) as potential sources of tritium supply. New analysis of the production capabilities of the Fast Flux Test Facility (FFTF) at the Hanford Site argues for considering its inclusion in the tritium supply,system. The use of the FFTF (alone or together with the Advanced Test Reactor [ATR] at the Idaho National Engineering Laboratory) as an integral part of,a tritium production system would help (1) ensure supply by 2005, (2) provide additional time to resolve institutional and technical issues associated with the- dual-track strategy, and (3) reduce discounted total life-cycle`costs and near-tenn annual expenditures for accelerator-based systems. The FFRF would also provide a way to get an early start.on dispositioning surplus weapons-usable plutonium as well as provide a source of medical isotopes. Challenges Associated With the Dual-Track Strategy The Departinent`s purchase of either a commercial reactor or reactor irradiation services faces challenging institutional issues associated with converting civilian reactors to defense uses. In addition, while the technical capabilities of the individual components of the accelerator have been proven, the entire system needs to be demonstrated and scaled upward to ensure that the components work toge ther 1548 as a complete production system. These challenges create uncertainty over the ability of the du2a-track strategy to provide an assured tritium supply source by 2005. Because the earliest the accelerator could come on line is 2007, it would have to operate at maximum capacity for the first few years to regenerate the reserves lost through radioactive decay aftei 2005.

  20. Tokamaks with high-performance resistive magnets: advanced test reactors and prospects for commercial applications

    SciTech Connect

    Bromberg, L.; Cohn, D.R.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

    1981-10-01

    Scoping studies have been made of tokamak reactors with high performance resistive magnets which maximize advantages gained from high field operation and reduced shielding requirements, and minimize resistive power requirements. High field operation can provide very high values of fusion power density and n tau/sub e/ while the resistive power losses can be kept relatively small. Relatively high values of Q' = Fusion Power/Magnet Resistive Power can be obtained. The use of high field also facilitates operation in the DD-DT advanced fuel mode. The general engineering and operational features of machines with high performance magnets are discussed. Illustrative parameters are given for advanced test reactors and for possible commercial reactors. Commercial applications that are discussed are the production of fissile fuel, electricity generation with and without fissioning blankets and synthetic fuel production.

  1. High Flux Isotope Reactor Core Analysis-Challenges and Recent Enhancements in Modeling and Simulation

    SciTech Connect

    Ilas, Germina

    2016-01-01

    A concerted effort over the past few years has focused on enhancing the core depletion models for the High Flux Isotope Reactor (HFIR) as part of a comprehensive study for designing a HFIR core that would use low-enriched uranium (LEU) fuel. A HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed for use as a reference for the design of an LEU fuel for HFIR and to improve the basis for analyses that support HFIR s current operation with high-enriched uranium (HEU) fuel. This paper summarizes the recent improvements in modeling and simulation for HFIR core analyses, with a focus on core depletion models.

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

    SciTech Connect

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

    1995-04-01

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

  3. Modeling and Depletion Simulations for a High Flux Isotope Reactor Cycle with a Representative Experiment Loading

    SciTech Connect

    Chandler, David; Betzler, Ben; Hirtz, Gregory John; Ilas, Germina; Sunny, Eva

    2016-09-01

    The purpose of this report is to document a high-fidelity VESTA/MCNP High Flux Isotope Reactor (HFIR) core model that features a new, representative experiment loading. This model, which represents the current, high-enriched uranium fuel core, will serve as a reference for low-enriched uranium conversion studies, safety-basis calculations, and other research activities. A new experiment loading model was developed to better represent current, typical experiment loadings, in comparison to the experiment loading included in the model for Cycle 400 (operated in 2004). The new experiment loading model for the flux trap target region includes full length 252Cf production targets, 75Se production capsules, 63Ni production capsules, a 188W production capsule, and various materials irradiation targets. Fully loaded 238Pu production targets are modeled in eleven vertical experiment facilities located in the beryllium reflector. Other changes compared to the Cycle 400 model are the high-fidelity modeling of the fuel element side plates and the material composition of the control elements. Results obtained from the depletion simulations with the new model are presented, with a focus on time-dependent isotopic composition of irradiated fuel and single cycle isotope production metrics.

  4. Fuel elements of research reactors in China

    SciTech Connect

    Yongmao, Z.; Dianshan, C.; Guofang, Q.

    1988-01-01

    This paper describes the current status of design, fabrication of fuel elements for research reactors in China, emphasis is placed on the technology of fuel elements for the High Flux Engineering Test Reactor (HFETR).

  5. Radiological characterization of the pressure vessel internals of the BNL High Flux Beam Reactor.

    PubMed

    Holden, Norman E; Reciniello, Richard N; Hu, Jih-Perng

    2004-08-01

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, measurements and calculations of the decay gamma-ray dose-rate were performed in the reactor pressure vessel and on vessel internal structures such as the upper and lower thermal shields, the Transition Plate, and the Control Rod blades. Measurements of gamma-ray dose rates were made using Red Perspex polymethyl methacrylate high-dose film, a Radcal "peanut" ion chamber, and Eberline's RO-7 high-range ion chamber. As a comparison, the Monte Carlo MCNP code and MicroShield code were used to model the gamma-ray transport and dose buildup. The gamma-ray dose rate at 8 cm above the center of the Transition Plate was measured to be 160 Gy h (using an RO-7) and 88 Gy h at 8 cm above and about 5 cm lateral to the Transition Plate (using Red Perspex film). This compares with a calculated dose rate of 172 Gy h using Micro-Shield. The gamma-ray dose rate was 16.2 Gy h measured at 76 cm from the reactor core (using the "peanut" ion chamber) and 16.3 Gy h at 87 cm from the core (using Red Perspex film). The similarity of dose rates measured with different instruments indicates that using different methods and instruments is acceptable if the measurement (and calculation) parameters are well defined. Different measurement techniques may be necessary due to constraints such as size restrictions.

  6. Application of amorphous filler metals in production of fusion reactor high heat flux components

    SciTech Connect

    Kalin, B.A.; Fedotov, V.T.; Grigoriev, A.E.

    1994-12-31

    The technology of Al-Si, Zr-Ti-Be and Ti-Zr-Cu-Ni amorphous filler metals for Be and graphite brazing with Cu, Mo and V was developed. The fusion reactor high heat flux components from Cu-Be, Cu-graphite, Mo-Be, Mo-graphite, V-Re and V-graphite materials were produced by brazing. Every component represents metallic base, to which Be or graphite plates are brazed. The distance between plates was equal 0.2 times the plate height. These components were irradiated by hydrogen plasma with 5 x 10{sup 6} W/m{sup 2} power. The microstructure and the element distribution in the brazed zone were investigated before and after heat plasma irradiation. Topography graphite plate surfaces and topography of metal surfaces between plates were also investigated after heat plasma irradiation. The results of microstructure investigation and material erosion are discussed.

  7. Structural biology facilities at Brookhaven National Laboratory`s high flux beam reactor

    SciTech Connect

    Korszun, Z.R.; Saxena, A.M.; Schneider, D.K.

    1994-12-31

    The techniques for determining the structure of biological molecules and larger biological assemblies depend on the extent of order in the particular system. At the High Flux Beam Reactor at the Brookhaven National Laboratory, the Biology Department operates three beam lines dedicated to biological structure studies. These beam lines span the resolution range from approximately 700{Angstrom} to approximately 1.5{Angstrom} and are designed to perform structural studies on a wide range of biological systems. Beam line H3A is dedicated to single crystal diffraction studies of macromolecules, while beam line H3B is designed to study diffraction from partially ordered systems such as biological membranes. Beam line H9B is located on the cold source and is designed for small angle scattering experiments on oligomeric biological systems.

  8. Production of Thorium-229 at the ORNL High Flux Isotope Reactor

    SciTech Connect

    Boll, Rose Ann; Garland, Marc A; Mirzadeh, Saed

    2008-01-01

    The investigation of targeted cancer therapy using -emitters has developed considerably in recent years and clinical trials have generated promising results. In particular, the initial clinical trials for treatment of acute myeloid leukemia have demonstrated the effectiveness of the -emitter 213Bi in killing cancer cells [1]. Pre-clinical studies have also shown the potential application of both 213Bi and its 225Ac parent radionuclide in a variety of cancer systems and targeted radiotherapy [2]. Bismuth-213 is obtained from a radionuclide generator system from decay of the 10-d 225Ac parent, a member of the 7340-y 229Th chain. Currently, 233U is the only viable source for high purity 229Th; however, due to increasing difficulties associated with 233U safeguards, processing additional 233U is presently unfeasible. The recent decision to downblend and dispose of enriched 233U further diminished the prospects for extracting 229Th from 233U stock. Nevertheless, the anticipated growth in demand for 225Ac may soon exceed the levels of 229Th (~40 g or ~8 Ci; ~80 times the current ORNL 229Th stock) present in the aged 233U stockpile. The alternative routes for the production of 229Th, 225Ra and 225Ac include both reactor and accelerator approaches [3]. Here, we describe production of 229Th via neutron transmutation of 226Ra targets in the ORNL High Flux Isotope Reactor (HFIR).

  9. Irradiation Facilities at the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2005-12-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC – formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world’s data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities1. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens.

  10. First annual progress report on United States-Japan Collaborative Testing in the High Flux Isotope Reactor and the Oak Ridge Research Reactor for the period ending September 30, 1984

    SciTech Connect

    Scott, J.L.; Grossbeck, M.L.; Jitsukaw, S.; Rowcliffe, A.F.; West, C.D.; Tanaka, M.P.; Kondo, T.; Hishinuma, A.

    1985-10-01

    The first year of the Program of US-Japan Collaborative Testing in HFIR and ORR has been completed. The program consists of eight Phase-I HFIR Target Capsules and two ORR Spectral-Tailoring Capsules equally shared. At the end of the first year, all eight HFIR capsules were completed and six were being irradiated in HFIR. Designs of both ORR Capsules were completed, and a thermal mock-up of the 60/200/sup 0/C capsule was successfully operated. Preparations were made for the installation of a new JEOL JEM-2000FX Microscope to be supplied by JAERI to ORNL for the collaboration. All work was completed on schedule and within costs.

  11. Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

    SciTech Connect

    Renfro, David G; Cook, David Howard; Freels, James D; Griffin, Frederick P; Ilas, Germina; Sease, John D; Chandler, David

    2012-03-01

    This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

  12. Feasibility analyses for HEU to LEU fuel conversion of the LAUE Langivin Institute (ILL) High Flux Reactor (RHF).

    SciTech Connect

    Stevens, J.; Tentner. A.; Bergeron, A.; Nuclear Engineering Division

    2010-08-19

    The High Flux Reactor (RHF) of the Laue Langevin Institute (ILL) based in Grenoble, France is a research reactor designed primarily for neutron beam experiments for fundamental science. It delivers one of the most intense neutron fluxes worldwide, with an unperturbed thermal neutron flux of 1.5 x 10{sup 15} n/cm{sup 2}/s in its reflector. The reactor has been conceived to operate at a nuclear power of 57 MW but currently operates at 52 MW. The reactor currently uses a Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most worldwide research and test reactors have already started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on a mixture of uranium and molybdenum (UMo) is expected to allow the conversion of compact high performance reactors like the RHF. This report presents the results of reactor design, performance and steady state safety analyses for conversion of the RHF from the use of HEU fuel to the use of UMo LEU fuel. The objective of this work was to show that is feasible, under a set of manufacturing assumptions, to design a new RHF fuel element that could safely replace the HEU element currently used. The new proposed design has been developed to maximize performance, minimize changes and preserve strong safety margins. Neutronics and thermal-hydraulics models of the RHF have been developed and qualified by benchmark against experiments and/or against other codes and models. The models developed were then used to evaluate the RHF performance if LEU UMo were to replace the current HEU fuel 'meat' without any geometric change to the fuel plates. Results of these direct replacement analyses have shown a significant degradation of the RHF performance, in terms of both neutron flux and cycle length

  13. Density of Gadolinium Nitrate Solutions for the High Flux Isotope Reactor

    SciTech Connect

    Taylor, Paul Allen; Lee, Denise L

    2009-05-01

    In late 1992, the High Flux Isotope Reactor (HFIR) was planning to switch the solution contained in the poison injection tank from cadmium nitrate to gadolinium nitrate. The poison injection system is an emergency system used to shut down the reactor by adding a neutron poison to the cooling water. This system must be able to supply a minimum of 69 pounds of gadolinium to the reactor coolant system in order to guarantee that the reactor would become subcritical. A graph of the density of gadolinium nitrate solutions over a concentration range of 5 to 30 wt% and a temperature range of 15 to 40{sup o}C was prepared. Routine density measurements of the solution in the poison injection tank are made by HFIR personnel, and an adaptation of the original graph is used to determine the gadolinium nitrate concentration. In late 2008, HFIR personnel decided that the heat tracing that was present on the piping for the poison injection system could be removed without any danger of freezing the solution; however, the gadolinium nitrate solution might get as cold as 5{sup o}C. This was outside the range of the current density-concentration correlation, so the range needed to be expanded. This report supplies a new density-concentration correlation that covers the extended temperature range. The correlation is given in new units, which greatly simplifies the calculation that is required to determine the pounds of gadolinium in the tank solution. The procedure for calculating the amount of gadolinium in the HFIR poison injection system is as follows: (1) Calculate the usable volume in the system; (2) Measure the density of the solution; (3) Calculate the gadolinium concentration using the following equation: Gd(lb/ft{sup 3}) = measured density (g/mL) x 34.681 - 34.785; (4) Calculate the amount of gadolinium in the system using the following equation: Amount of Gd(lb) = Gd concentration (lb/ft{sup 3}) x usable volume (ft{sup 3}). The equation in step 3 is exact for a temperature of

  14. Engineered Coatings for Ni Alloys in High Temperature Reactors

    NASA Astrophysics Data System (ADS)

    Clark, Elizabeth A.; Yang, James Y.; Kumar, Deepak; Was, Gary S.; Levi, Carlos G.

    2013-02-01

    Alloy 617 is a candidate material for the intermediate heat exchanger of the He-cooled very high temperature reactor. At target temperatures ≥1223 K (950 °C), low level impurities in the gas stream may cause carburization, decarburization, and/or oxidation of 617 with deleterious effects on its mechanical properties. The chromia scale formed naturally by 617 does not provide adequate protection in the expected environment. Alpha alumina offers a greater potential as an effective diffusion barrier with superior stability, but it requires modification of the alloy surface. This work explores two approaches to surface modification based on aluminizing, either alone or in combination with FeCrAlY cladding, followed by pre-oxidation to form alpha alumina. Both approaches yield coatings with promising diffusional stability on alloy 617. Initial corrosion studies in impure He environments reveal that the alpha alumina is stable and protects the underlying substrate in both carburizing and decarburizing environments.

  15. Job/task analysis for I C (Instrumentation and Controls) instrument technicians at the High Flux Isotope Reactor

    SciTech Connect

    Duke, L.L.

    1989-09-01

    To comply with Department of Energy Order 5480.XX (Draft), a job/task analysis was initiated by the Maintenance Management Department at Oak Ridge National Laboratory (ORNL). The analysis was applicable to instrument technicians working at the ORNL High Flux Isotope Reactor (HFIR). This document presents the procedures and results of that analysis. 2 refs., 2 figs.

  16. High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

    SciTech Connect

    Donna Post Guillen; Larry R. Greenwood; James R. Parry

    2014-06-22

    A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

  17. Very high temperature measurements: Application to nuclear reactor safety tests

    NASA Astrophysics Data System (ADS)

    Parga, Clemente Jose

    This PhD dissertation focuses on the improvement of very high temperature thermometry (1100ºC to 2480ºC), with special emphasis on the application to the field of nuclear reactor safety and severe accident research. Two main projects were undertaken to achieve this objective: -The development, testing and transposition of high-temperature fixed point (HTFP) metal-carbon eutectic cells, from metrology laboratory precision (+/-0.001ºC) to applied research with a reasonable degradation of uncertainties (+/-3-5ºC). -The corrosion study and metallurgical characterization of Type-C thermocouple (service temp. 2300ºC) prospective sheath material was undertaken to extend the survivability of TCs used for molten metallic/oxide corium thermometry (below 2000ºC).

  18. Modeling and Simulations for the High Flux Isotope Reactor Cycle 400

    SciTech Connect

    Ilas, Germina; Chandler, David; Ade, Brian J; Sunny, Eva E; Betzler, Benjamin R; Pinkston, Daniel

    2015-03-01

    A concerted effort over the past few years has been focused on enhancing the core model for the High Flux Isotope Reactor (HFIR), as part of a comprehensive study for HFIR conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel. At this time, the core model used to perform analyses in support of HFIR operation is an MCNP model for the beginning of Cycle 400, which was documented in detail in a 2005 technical report. A HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed to serve as reference for the design of an LEU fuel for HFIR. The recent enhancements in modeling and simulations for HFIR that are discussed in the present report include: (1) revision of the 2005 MCNP model for the beginning of Cycle 400 to improve the modeling data and assumptions as necessary based on appropriate primary reference sources HFIR drawings and reports; (2) improvement of the fuel region model, including an explicit representation for the involute fuel plate geometry that is characteristic to HFIR fuel; and (3) revision of the Monte Carlo-based depletion model for HFIR in use since 2009 but never documented in detail, with the development of a new depletion model for the HFIR explicit fuel plate representation. The new HFIR models for Cycle 400 are used to determine various metrics of relevance to reactor performance and safety assessments. The calculated metrics are compared, where possible, with measurement data from preconstruction critical experiments at HFIR, data included in the current HFIR safety analysis report, and/or data from previous calculations performed with different methods or codes. The results of the analyses show that the models presented in this report provide a robust and reliable basis for HFIR analyses.

  19. Utilization of the High Flux Isotope Reactor at Oak Ridge National Laboratory

    SciTech Connect

    Selby, Douglas L; Bilheux, Hassina Z; Meilleur, Flora; Jones, Amy; Bailey, William Barton; Vandergriff, David H

    2015-01-01

    This paper addresses several aspects of the scientific utilization of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR). Topics to be covered will include: 1) HFIR neutron scattering instruments and the formal instrument user program; 2) Recent upgrades to the neutron scattering instrument stations at the reactor, and 3) eMod a new tool for addressing instrument modifications and providing configuration control and design process for scientific instruments at HFIR and the Spallation Neutron Source (SNS). There are 15 operating neutron instrument stations at HFIR with 12 of them organized into a formal user program. Since the last presentation on HFIR instruments at IGORR we have installed a Single Crystal Quasi-Laue Diffractometer instrument called IMAGINE; and we have made significant upgrades to HFIR neutron scattering instruments including the Cold Triple Axis Instrument, the Wide Angle Neutron Diffractometer, the Powder Diffractometer, and the Neutron Imaging station. In addition, we have initiated upgrades to the Thermal Triple Axis Instrument and the Bio-SANS cold neutron instrument detector system. All of these upgrades are tied to a continuous effort to maintain a high level neutron scattering user program at the HFIR. For the purpose of tracking modifications such as those mentioned and configuration control we have been developing an electronic system for entering instrument modification requests that follows a modification or instrument project through concept development, design, fabrication, installation, and commissioning. This system, which we call eMod, electronically leads the task leader through a series of questions and checklists that then identifies such things as ES&H and radiological issues and then automatically designates specific individuals for the activity review process. The system has been in use for less than a year and we are still working out some of the inefficiencies, but we believe that this will become a very

  20. Reactivity Accountability Attributed to Reflector Poisons in the High Flux Isotope Reactor

    SciTech Connect

    Chandler, David; Maldonado, G Ivan; Primm, Trent

    2009-12-01

    The objective of this study is to develop a methodology to predict the reactivity impact as a function of outage time between cycles of 3He, 6Li, and other poisons in the High Flux Isotope Reactor s (HFIR) beryllium reflector. The reactivity worth at startup of the HFIR has been incorrectly predicted in the past after the reactor has been shut-down for long periods of time. The incorrect prediction was postulated to be due to the erroneous calculation of 3He buildup in the beryllium reflector. It is necessary to develop a better estimate of the start-of-cycle symmetric critical control element positions since if the estimated and actual symmetrical critical control element positions differ by more than $1.55 in reactivity (approximately one-half inch in control element startup position), HFIR is to be shutdown and a technical evaluation is performed to resolve the discrepancy prior to restart. 3He is generated and depleted during operation, but during an outage, the depletion of 3He ceases because it is a stable isotope. 3He is born from the radioactive decay of tritium, and thus the concentration of 3He increases during shutdown. SCALE, specifically the TRITON and CSAS5 control modules including the KENO V.A, COUPLE, and ORIGEN functional modules were utilized in this study. An equation relating the down time (td) to the change in symmetric control element position was generated and validated against measurements for approximately 40 HFIR operating cycles. The newly-derived correlation was shown to improve accuracy of predictions for long periods of down time.

  1. Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling

    SciTech Connect

    Travis, Adam R

    2014-05-01

    A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.

  2. Awareness, Preference, Utilization, and Messaging Research for the Spallation Neutron Source and High Flux Isotope Reactor

    SciTech Connect

    Bryant, Rebecca; Kszos, Lynn A

    2011-03-01

    Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one-on-one interviews

  3. Proposals for ORNL (Oak Ridge National Laboratory) support to Tiber LLNL (Lawrence Livermore National Laboratory). [Engineering Test Reactor

    SciTech Connect

    Berry, L.A.; Rosenthal, M.W.; Saltmarsh, M.J.; Shannon, T.E.; Sheffield, J.

    1987-01-27

    This document describes the interests and capabilities of Oak Ridge National Laboratory in their proposals to support the Lawrence Livermore National Laboratory (LLNL) Engineering Test Reactor (ETR) project. Five individual proposals are cataloged separately. (FI)

  4. Tritium trapping in silicon carbide in contact with solid breeder under high flux isotope reactor irradiation

    SciTech Connect

    H. Katsui; Y. Katoh; A. Hasegawa; M. Shimada; Y. Hatano; T. Hinoki; S. Nogami; T. Tanaka; S. Nagata; T. Shikama

    2013-11-01

    The trapping of tritium in silicon carbide (SiC) injected from ceramic breeding materials was examined via tritium measurements using imaging plate (IP) techniques. Monolithic SiC in contact with ternary lithium oxide (lithium titanate and lithium aluminate) as a ceramic breeder was irradiated in the High Flux Isotope Reactor (HFIR) in Oak Ridge, Tennessee, USA. The distribution of photo-stimulated luminescence (PSL) of tritium in SiC was successfully obtained, which separated the contribution of 14C ß-rays to the PSL. The tritium incident from ceramic breeders was retained in the vicinity of the SiC surface even after irradiation at 1073 K over the duration of ~3000 h, while trapping of tritium was not observed in the bulk region. The PSL intensity near the SiC surface in contact with lithium titanate was higher than that obtained with lithium aluminate. The amount of the incident tritium and/or the formation of a Li2SiO3 phase on SiC due to the reaction with lithium aluminate under irradiation likely were responsible for this observation.

  5. Experimental and analytical studies of high heat flux components for fusion experimental reactor

    NASA Astrophysics Data System (ADS)

    Araki, Masanori

    1993-03-01

    In this report, the experimental and analytical results concerning the development of plasma facing components of ITER are described. With respect to developing high heat removal structures for the divertor plates, an externally-finned swirl tube was developed based on the results of critical heat flux (CHF) experiments on various tube structures. As the result, the burnout heat flux, which also indicates incident CHF, of 41 (+/-) 1 MW/sq m was achieved in the externally-finned swirl tube. The applicability of existing CHF correlations based on uniform heating conditions was evaluated by comparing the CHF experimental data with the smooth and the externally-finned tubes under one-sided heating condition. As the results, experimentally determined CHF data for straight tube show good agreement, for the externally-finned tube, no existing correlations are available for prediction of the CHF. With respect to the evaluation of the bonds between carbon-based material and heat sink metal, results of brazing tests were compared with the analytical results by three dimensional model with temperature-dependent thermal and mechanical properties. Analytical results showed that residual stresses from brazing can be estimated by the analytical three directional stress values instead of the equivalent stress value applied. In the analytical study on the separatrix sweeping for effectively reducing surface heat fluxes on the divertor plate, thermal response of the divertor plate was analyzed under ITER relevant heat flux conditions and has been tested. As the result, it has been demonstrated that application of the sweeping technique is very effective for improvement in the power handling capability of the divertor plate and that the divertor mock-up has withstood a large number of additional cyclic heat loads.

  6. Deterministic Modeling of the High Temperature Test Reactor

    SciTech Connect

    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

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

    SciTech Connect

    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).

  8. Development of Advanced Thermal and Environmental Barrier Coatings Using a High-Heat-Flux Testing Approach

    NASA Technical Reports Server (NTRS)

    Zhu, Dongming; Miller, Robert A.

    2003-01-01

    The development of low conductivity, robust thermal and environmental barrier coatings requires advanced testing techniques that can accurately and effectively evaluate coating thermal conductivity and cyclic resistance at very high surface temperatures (up to 1700 C) under large thermal gradients. In this study, a laser high-heat-flux test approach is established for evaluating advanced low conductivity, high temperature capability thermal and environmental barrier coatings under the NASA Ultra Efficient Engine Technology (UEET) program. The test approach emphasizes the real-time monitoring and assessment of the coating thermal conductivity, which initially rises under the steady-state high temperature thermal gradient test due to coating sintering, and later drops under the cyclic thermal gradient test due to coating cracking/delamination. The coating system is then evaluated based on damage accumulation and failure after the combined steady-state and cyclic thermal gradient tests. The lattice and radiation thermal conductivity of advanced ceramic coatings can also be evaluated using laser heat-flux techniques. The external radiation resistance of the coating is assessed based on the measured specimen temperature response under a laser- heated intense radiation-flux source. The coating internal radiation contribution is investigated based on the measured apparent coating conductivity increases with the coating surface test temperature under large thermal gradient test conditions. Since an increased radiation contribution is observed at these very high surface test temperatures, by varying the laser heat-flux and coating average test temperature, the complex relation between the lattice and radiation conductivity as a function of surface and interface test temperature may be derived.

  9. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

    SciTech Connect

    Primm, Trent; Chandler, David; Ilas, Germina; Miller, James Henry; Sease, John D; Jolly, Brian C

    2009-03-01

    This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

  10. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual report for FY 2009

    SciTech Connect

    Chandler, David; Freels, James D; Ilas, Germina; Miller, James Henry; Primm, Trent; Sease, John D; Guida, Tracey; Jolly, Brian C

    2010-02-01

    This report documents progress made during FY 2009 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Studies are reported of the application of a silicon coating to surrogates for spheres of uranium-molybdenum alloy. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. A description of the progress in developing a finite element thermal hydraulics model of the LEU core is provided.

  11. Development of CFD models to support LEU Conversion of ORNL s High Flux Isotope Reactor

    SciTech Connect

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

    2012-01-01

    The US Department of Energy s National Nuclear Security Administration (NNSA) is participating in the Global Threat Reduction Initiative to reduce and protect vulnerable nuclear and radiological materials located at civilian sites worldwide. As an integral part of one of NNSA s subprograms, Reduced Enrichment for Research and Test Reactors, HFIR is being converted from the present HEU core to a low enriched uranium (LEU) core with less than 20% of U-235 by weight. Because of HFIR s importance for condensed matter research in the United States, its conversion to a high-density, U-Mo-based, LEU fuel should not significantly impact its existing performance. Furthermore, cost and availability considerations suggest making only minimal changes to the overall HFIR facility. Therefore, the goal of this conversion program is only to substitute LEU for the fuel type in the existing fuel plate design, retaining the same number of fuel plates, with the same physical dimensions, as in the current HFIR HEU core. Because LEU-specific testing and experiments will be limited, COMSOL Multiphysics was chosen to provide the needed simulation capability to validate against the HEU design data and previous calculations, and predict the performance of the proposed LEU fuel for design and safety analyses. To achieve it, advanced COMSOL-based multiphysics simulations, including computational fluid dynamics (CFD), are being developed to capture the turbulent flows and associated heat transfer in fine detail and to improve predictive accuracy [2].

  12. Polymerization Reactor Engineering.

    ERIC Educational Resources Information Center

    Skaates, J. Michael

    1987-01-01

    Describes a polymerization reactor engineering course offered at Michigan Technological University which focuses on the design and operation of industrial polymerization reactors to achieve a desired degree of polymerization and molecular weight distribution. Provides a list of the course topics and assigned readings. (TW)

  13. Conceptual Process for the Manufacture of Low-Enriched Uranium/Molybdenum Fuel for the High Flux Isotope Reactor

    SciTech Connect

    Sease, J.D.; Primm, R.T. III; Miller, J.H.

    2007-09-30

    The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are within the

  14. Thermal Safety Analyses for the Production of Plutonium-238 at the High Flux Isotope Reactor

    SciTech Connect

    Hurt, Christopher J.; Freels, James D.; Hobbs, Randy W.; Jain, Prashant K.; Maldonado, G. Ivan

    2016-08-01

    There has been a considerable effort over the previous few years to demonstrate and optimize the production of plutonium-238 (238Pu) at the High Flux Isotope Reactor (HFIR). This effort has involved resources from multiple divisions and facilities at the Oak Ridge National Laboratory (ORNL) to demonstrate the fabrication, irradiation, and chemical processing of targets containing neptunium-237 (237Np) dioxide (NpO2)/aluminum (Al) cermet pellets. A critical preliminary step to irradiation at the HFIR is to demonstrate the safety of the target under irradiation via documented experiment safety analyses. The steady-state thermal safety analyses of the target are simulated in a finite element model with the COMSOL Multiphysics code that determines, among other crucial parameters, the limiting maximum temperature in the target. Safety analysis efforts for this model discussed in the present report include: (1) initial modeling of single and reduced-length pellet capsules in order to generate an experimental knowledge base that incorporate initial non-linear contact heat transfer and fission gas equations, (2) modeling efforts for prototypical designs of partially loaded and fully loaded targets using limited available knowledge of fabrication and irradiation characteristics, and (3) the most recent and comprehensive modeling effort of a fully coupled thermo-mechanical approach over the entire fully loaded target domain incorporating burn-up dependent irradiation behavior and measured target and pellet properties, hereafter referred to as the production model. These models are used to conservatively determine several important steady-state parameters including target stresses and temperatures, the limiting condition of which is the maximum temperature with respect to the melting point. The single pellet model results provide a basis for the safety of the irradiations, followed by parametric analyses in the initial prototypical designs that were necessary due to the

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

    SciTech Connect

    Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville; Gougar, Hans David; Strydom, Gerhard

    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.

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

    SciTech Connect

    Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville; Gougar, Hans David; Kinsey, J.; Strydom, Gerhard

    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.

  17. Failure analysis of beryllium tile assembles following high heat flux testing for the ITER program

    SciTech Connect

    B. C. Odegard, Jr.; C. H. Cadden; N. Y. C. Yang

    2000-05-01

    The following document describes the processing, testing and post-test analysis of two Be-Cu assemblies that have successfully met the heat load requirements for the first wall and dome sections for the ITER (International Thermonuclear Experimental Reactor) fusion reactor. Several different joint assemblies were evaluated in support of a manufacturing technology investigation aimed at diffusion bonding or brazing a beryllium armor tile to a copper alloy heat sink for fusion reactor applications. Judicious selection of materials and coatings for these assemblies was essential to eliminate or minimize interactions with the highly reactive beryllium armor material. A thin titanium layer was used as a diffusion barrier to isolate the copper heat sink from the beryllium armor. To reduce residual stresses produced by differences in the expansion coefficients between the beryllium and copper, a compliant layer of aluminum or aluminum-beryllium (AlBeMet-150) was used. Aluminum was chosen because it does not chemically react with, and exhibits limited volubility in, beryllium. Two bonding processes were used to produce the assemblies. The primary process was a diffusion bonding technique. In this case, undesirable metallurgical reactions were minimized by keeping the materials in a solid state throughout the fabrication cycle. The other process employed an aluminum-silicon layer as a brazing filler material. In both cases, a hot isostatic press (HIP) furnace was used in conjunction with vacuum-canned assemblies in order to minimize oxidation and provide sufficient pressure on the assemblies for full metal-to-metal contact and subsequent bonding. The two final assemblies were subjected to a suite of tests including: tensile tests and electron and optical metallography. Finally, high heat flux testing was conducted at the electron beam testing system (EBTS) at Sandia National Laboratories, New Mexico. Here, test mockups were fabricated and subjected to normal heat loads to

  18. Summary of HTGR (high-temperature gas-cooled reactor) benchmark data from the high temperature lattice test reactor

    SciTech Connect

    Newman, D.F.

    1989-10-01

    The High Temperature Lattice Test Reactor (HTLTR) was a unique critical facility specifically built and operated to measure variations in neutronic characteristics of high temperature gas cooled reactor (HTGR) lattices at temperatures up to 1000{degree}C. The Los Alamos National Laboratory commissioned Pacific Northwest Laboratory (PNL) to prepare this summary reference report on the HTLTR benchmark data and its associated documentation. In the initial stages of the program, the principle of the measurement of k{sub {infinity}} using the unpoisoned technique (developed by R.E. Heineman of PNL) was subjected to extensive peer review within PNL and the General Atomic Company. A number of experiments were conducted at PNL in the Physical Constants Testing Reactor (PCTR) using both the unpoisoned technique and the well-established null reactivity technique that substantiated the equivalence of the measurements by direct comparison. Records of all data from fuel fabrication, the reactor experiments, and the analytical results were compiled and maintained to meet applicable quality assurance standards in place at PNL. Sensitivity of comparisons between measured and calculated k{sub {infinity}}(T) data for various HTGR lattices to changes in neutron cross section data, graphite scattering kernel models, and fuel block loading variations, were analyzed by PNL for the Electric Power Research Institute. As a part of this effort, the fuel rod composition in the dilute {sup 233}UO{sub 2}-ThO{sub 2} HTGR central cell (HTLTR Lattice {number sign}3) was sampled and analyzed by mass spectrometry. Values of k{sub {infinity}} calculated for that lattice were about 5% higher than those measured. Trace quantities of sodium chloride were found in the fuel rod that were equivalent to 22 atom parts-per-million of natural boron.

  19. Development of Low Conductivity and Ultra High Temperature Ceramic Coatings Using A High-Heat-Flux Testing Approach

    NASA Technical Reports Server (NTRS)

    Zhu, Dongming; Miller, Robert A.

    1990-01-01

    The development of low conductivity, robust thermal and environmental barrier coatings requires advanced testing techniques that can accurately and effectively evaluate coating thermal conductivity and cyclic resistance at very high surface temperatures (up to 17OOOC) under large thermal gradients. In this study, a laser high-heat-flux test approach is established for evaluating advanced low conductivity, ultra-high temperature ceramic thermal and environmental barrier coatings under the NASA Ultra Efficient Engine Technology (UEET) program. The test approach emphasizes the real-time monitoring and assessment of the coating thermal conductivity: the initial conductivity rise under a steady-state high temperature thermal gradient test due to coating sintering, and the later coating conductivity reduction under a subsequent cyclic thermal gradient test due to coating cracking/delamination. The coating system is then evaluated based on the damage accumulations and failure after the combined steady-state and cyclic thermal gradient tests. The lattice and radiation thermal conductivity of advanced ceramic coatings can also be evaluated using laser heat-flux techniques. The coating external radiation resistance is assessed based on the measured specimen temperature response under a laser heated intense radiation flux source. The coating internal radiation contribution is investigated based on the measured apparent coating conductivity increases with the coating surface test temperature under large thermal gradient test conditions. Since an increased radiation contribution is observed at these very high surface test temperatures, by varying the laser heat-flux and coating average test temperature, the complex relation between the lattice and radiation conductivity as a function of surface and interface test temperature is derived.

  20. A Project Management and Systems Engineering Structure for a Generation IV Very High Temperature Reactor

    SciTech Connect

    Ed Gorski; Dennis Harrell; Finis Southworth

    2004-09-01

    The Very High Temperature Reactor (VHTR) will be an advanced, very high temperature (approximately 1000o C. coolant outlet temperature), gas cooled nuclear reactor and is the nearest term of six Generation IV reactor technologies for nuclear assisted hydrogen production. In 2001, the Generation IV International Forum (GIF), a ten nation international forum working together with the Department of Energy’s (DOE) Nuclear Energy Research Advisory Committee (NERAC), agreed to proceed with the development of a technology roadmap and identified the next generation of nuclear reactor systems for producing new sources of power. Since a new reactor has not been licensed in the United States since the 1970s, the risks are too large for a single utility to assume in the development of an unprecedented Generation IV reactor. The government must sponsor and invest in the research to resolve major first of a kind (FOAK) issues through a full-scale demonstration prior to industry implementation. DOE’s primary mission for the VHTR is to demonstrate nuclear reactor assisted cogeneration of electricity and hydrogen while meeting the Generation IV goals for safety, sustainability, proliferation resistance and physical security and economics. The successful deployment of the VHTR as a demonstration project will aid in restarting the now atrophied U.S. nuclear power industry infrastructure. It is envisioned that VHTR project participants will include DOE Laboratories, industry partners such as designers, constructors, manufacturers, utilities, and Generation IV international countries. To effectively mange R&D, engineering, procurement, construction, and operation for this multi-organizational and technologically complex project, systems engineering will be used extensively to ensure delivery of the final product. Although the VHTR is an unprecedented FOAK system, the R&D, when assessed using the Office of Science and Technology Gate Model, falls primarily in the 3rd - Exploratory

  1. Multiphysics Simulations of the Complex 3D Geometry of the High Flux Isotope Reactor Fuel Elements Using COMSOL

    SciTech Connect

    Freels, James D; Jain, Prashant K

    2011-01-01

    A research and development project is ongoing to convert the currently operating High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory (ORNL) from highly-enriched Uranium (HEU U3O8) fuel to low-enriched Uranium (LEU U-10Mo) fuel. Because LEU HFIR-specific testing and experiments will be limited, COMSOL is chosen to provide the needed multiphysics simulation capability to validate against the HEU design data and calculations, and predict the performance of the LEU fuel for design and safety analyses. The focus of this paper is on the unique issues associated with COMSOL modeling of the 3D geometry, meshing, and solution of the HFIR fuel plate and assembled fuel elements. Two parallel paths of 3D model development are underway. The first path follows the traditional route through examination of all flow and heat transfer details using the Low-Reynolds number k-e turbulence model provided by COMSOL v4.2. The second path simplifies the fluid channel modeling by taking advantage of the wealth of knowledge provided by decades of design and safety analyses, data from experiments and tests, and HFIR operation. By simplifying the fluid channel, a significant level of complexity and computer resource requirements are reduced, while also expanding the level and type of analysis that can be performed with COMSOL. Comparison and confirmation of validity of the first (detailed) and second (simplified) 3D modeling paths with each other, and with available data, will enable an expanded level of analysis. The detailed model will be used to analyze hot-spots and other micro fuel behavior events. The simplified model will be used to analyze events such as routine heat-up and expansion of the entire fuel element, and flow blockage. Preliminary, coarse-mesh model results of the detailed individual fuel plate are presented. Examples of the solution for an entire fuel element consisting of multiple individual fuel plates produced by the simplified model are also presented.

  2. Fusion power demonstration - a baseline for the mirror engineering test reactor

    SciTech Connect

    Henning, C.D.; Logan, B.G.; Neef, W.S.; Dorn, D.; Clarkson, I.R.; Carpenter, T.; Gordon, J.D.; Campbell, R.B.; Hsu, P.; Nelson, D.

    1983-12-02

    Developing a definition of an engineering test reactor (ETR) is a current goal of the Office of Fusion Energy (OFE). As a baseline for the mirror ETR, the Fusion Power Demonstration (FPD) concept has been pursued at Lawrence Livermore National Laboratory (LLNL) in cooperation with Grumman Aerospace, TRW, and the Idaho National Engineering Laboratory. Envisioned as an intermediate step to fusion power applications, the FPD would achieve DT ignition in the central cell, after which blankets and power conversion would be added to produce net power. To achieve ignition, a minimum central cell length of 67.5 m is needed to supply the ion and alpha particles radial drift pumping losses in the transition region. The resulting fusion power is 360 MW. Low electron-cyclotron heating power of 12 MW, ion-cyclotron heating of 2.5 MW, and a sloshing ion beam power of 1.0 MW result in a net plasma Q of 22. A primary technological challenge is the 24-T, 45-cm bore choke coil, comprising a copper hybrid insert within a 15 to 18 T superconducting coil.

  3. Heat Flux Sensor Testing

    NASA Astrophysics Data System (ADS)

    Clark, D. W.

    2002-07-01

    This viewgraph presentation provides information on the following objectives: Developing secondary calibration capabilities for MSFC's (Marshall Space Flight Center) Hot Gas Facility (HGF), a Mach 4 Aerothermal Wind Tunnel; Evaluating ASTM (American Society for Testing and Materials) slug/ thinskin calorimeters against current HGF heat flux sensors; Providing verification of baselined AEDC (Arnold Engineering Development Center) / Medtherm gage calibrations; Addressing future calibration issues involving NIST (National Institute of Standards and Technology) certified radiant gages.

  4. Heat Flux Sensor Testing

    NASA Technical Reports Server (NTRS)

    Clark, D. W.

    2002-01-01

    This viewgraph presentation provides information on the following objectives: Developing secondary calibration capabilities for MSFC's (Marshall Space Flight Center) Hot Gas Facility (HGF), a Mach 4 Aerothermal Wind Tunnel; Evaluating ASTM (American Society for Testing and Materials) slug/ thinskin calorimeters against current HGF heat flux sensors; Providing verification of baselined AEDC (Arnold Engineering Development Center) / Medtherm gage calibrations; Addressing future calibration issues involving NIST (National Institute of Standards and Technology) certified radiant gages.

  5. Engineering Evaluation/Cost Analysis for Decommissioning of the Engineering Test Reactor Complex

    SciTech Connect

    A. B. Culp

    2006-10-01

    Preparation of this Engineering Evaluation/Cost Analysis is consistent with the joint U.S. Department of Energy and U.S. Environmental Protection Agency Policy on Decommissioning of Department of Energy Facilities Under the Comprehensive Environmental Response, Compensation, and Liability Act, which establishes the Comprehensive Environmental Response, Compensation, and Liability Act non-time-critical removal action (NTCRA) process as an approach for decommissioning.

  6. Miniaturized High Speed Controls for Turbine Engines (Fabrication and Test)

    DTIC Science & Technology

    1974-08-01

    AO/A-006 108 MINIATURIZED HIGH SPEED CONTROLS FOR TURBINE ENGINES (FABRICATION AND TEST ) D. G. Burnell, et al Colt lndustries, Incorporated Prepared...Speed Controlsma193-Ag97 for Turbine Engines (Fabrication and May 1973RIN ORD REugR 1974 Test ) 6.PRFRIA GOG EOTNME 7. AUTHOR(.) 6- CONTRACT OR GRANT... y asd Id..,tify by block numnb.) ’-This report summarizes the design and development of con- trol components and high speed fuel pump technology for

  7. Neutronics Conversion Analyses of the Laue-Langevin Institute (ILL) High Flux Reactor (RHF)

    SciTech Connect

    Bergeron, A.; Dionne, B.; Calzavara, Y.

    2014-09-30

    The following report describes the neutronics results obtained with the MCNP model of the RHF U7Mo LEU reference design that has been established in 2010 during the feasibility analysis. This work constitutes a complete and detailed neutronics analysis of that LEU design using models that have been significantly improved since 2010 and the release of the feasibility report. When possible, the credibility of the neutronics model is tested by comparing the HEU model results with experimental data or other codes calculations results. The results obtained with the LEU model are systematically compared to the HEU model. The changes applied to the neutronics model lead to better comparisons with experimental data or improved the calculation efficiency but do not challenge the conclusion of the feasibility analysis. If the U7Mo fuel is commercially available, not cost prohibitive, a back-end solution is established and if it is possible to manufacture the proposed element, neutronics analyses show that the performance of the reactor would not be challenged by the conversion to LEU fuel.

  8. Manufacturing and thermomechanical testing of actively cooled all beryllium high heat flux test pieces

    SciTech Connect

    Vasiliev, N.N.; Sokolov, Yu.A.; Shatalov, G.E.

    1995-09-01

    One of the problems affiliated to ITER high heat flux elements development is a problem of interface of beryllium protection with heat sink routinely made of copper alloys. To get rid of this problem all beryllium elements could be used as heat receivers in places of enhanced thermal loads. In accordance with this objectives four beryllium test pieces of two types have been manufactured in {open_quotes}Institute of Beryllium{close_quotes} for succeeding thermomechanical testing. Two of them were manufactured in accordance with JET team design; they are round {open_quotes}hypervapotron type{close_quotes} test pieces. Another two ones are rectangular test sections with a twisted tape installed inside of the circular channel. Preliminary stress-strain analysis have been performed for both type of the test pieces. Hypervapotrons have been shipped to JET where they were tested on JET test bed. Thermomechanical testing of pieces of the type of {open_quotes}swirl tape inside of tube{close_quotes} have been performed on Kurchatov Institute test bed. Chosen beryllium grade properties, some details of manufacturing, results of preliminary stress-strain analysis and thermomechanical testing of the test pieces {open_quotes}swirl tape inside of tube{close_quotes} type are given in this report.

  9. Flux effect analysis in WWER-440 reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Kryukov, A.; Blagoeva, D.; Debarberis, L.

    2013-11-01

    The results of long term research programme concerning the determination of irradiation embrittlement dependence on fast neutron flux for WWER-440 reactor pressure vessel steels before and after annealing are presented in this paper. The study of flux effect was carried out on commercial WWER-440 steels which differ significantly in phosphorous (0.013-0.036 wt%) and copper (0.08-0.20 wt%) contents. All specimens were irradiated in surveillance channel positions under similar conditions at high ˜4 × 1012 сm-2 s-1 and low ˜6 × 1011 сm-2 s-1 fluxes (E > 0.5 MeV) at a temperature of 270 °С. The radiation embrittlement was evaluated by transition temperature shift on the basis of Charpy specimens test results. In case of low flux, the measured Tk shifts could be 25-50 °C bigger than the Tk shifts obtained from high flux data. A significant flux effect is observed in WWER-440 reactor pressure vessel steels with higher copper content (>0.13 wt%).

  10. Advanced Test Reactor National Scientific User Facility

    SciTech Connect

    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.

  11. Preliminary requirements for a Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR)

    SciTech Connect

    Massie, M.; Forsberg, C.; Forget, B.; Hu, L. W.

    2012-07-01

    A Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR) design is being developed at MIT to provide the first demonstration and test of a salt-cooled reactor using high-temperature fuel. The first step is to define the requirements. The top level requirements are (1) provide the confidence that a larger demonstration reactor is warranted and (2) develop the necessary data for a larger-scale reactor. Because requirements will drive the design of the FHTR, a significant effort is being undertaken to define requirements and understand the tradeoffs that will be required for a practical design. The preliminary requirements include specifications for design parameters and necessary tests of major reactor systems. Testing requirements include demonstration of components, systems, and procedures for refueling, instrumentation, salt temperature control to avoid coolant freezing, salt chemistry and volume control, tritium monitoring and control, and in-service inspection. Safety tests include thermal hydraulics, neutronics - including intrinsic core shutdown mechanisms such as Doppler feedback - and decay heat removal systems. Materials and coolant testing includes fuels (including mechanical wear and fatigue) and system corrosion behavior. Preliminary analysis indicates a thermal power output below 30 MW, an initial core using pebble-bed or prismatic-block fuel, peak outlet temperatures of at least 700 deg. C, and use of FLi{sup 7}Be ({sup 7}LiF-BeF{sub 2}) coolant. The option to change-out the reactor core, fuel type, and major components is being investigated. While the FHTR will be used for materials testing, its primary mission is as a reactor system performance test to enable the design and licensing of a FHR demonstration power reactor. (authors)

  12. High-flux, high-temperature thermal vacuum qualification testing of a solar receiver aperture shield

    SciTech Connect

    Kerslake, T.W.; Mason, L.S.; Strumpf, H.J.

    1997-12-31

    As part of the International Space Station (ISS) Phase 1 program, NASA Lewis Research Center (LeRC) and the Russian Space Agency (RSA) teamed together to design, build and flight test the world`s first orbital Solar Dynamic Power System (SDPS) on the Russian space station Mir. The Solar Dynamic Flight Demonstration (SDFD) program was to operate a nominal 2 kWe SDPS on Mir for a period up to 1-year starting in late 1997. Unfortunately, the SDFD mission was demanifested from the ISS Phase 1 shuttle program in early 1996. However, substantial flight hardware and prototypical flight hardware was built including a heat receiver and aperture shield. The aperture shield comprises the front face of the cylindrical cavity heat receiver and is located at the focal plane of the solar concentrator. It is constructed of a stainless steel plate with a 1-m outside diameter, a 0.24-m inside diameter and covered with high-temperature, refractory metal multi-foil insulation (MFI). The aperture shield must minimize heat loss from the receiver cavity, provide a stiff, high strength structure to accommodate shuttle launch loads and protect receiver structures from highly concentrated solar fluxes during concentrator off-pointing events. To satisfy Mir operational safety protocols, the aperture shield was required to accommodate direct impingement of the intensely concentrated solar image for a 1-hour period. To verify thermal-structural durability under the anticipated high-flux, high-temperature loading, an aperture shield test article was constructed and underwent a series of two tests in a large thermal vacuum chamber configured with a reflective, point-focus solar concentrator and a solar simulator. The test article was positioned near the focal plane and exposed to concentrated solar flux for a period of 1-hour. In the first test, a near equilibrium temperature of 1862 K was attained in the center of the shield hot spot. In the second test, with increased incident flux, a near

  13. High-Flux, High-Temperature Thermal Vacuum Qualification Testing of a Solar Receiver Aperture Shield

    NASA Technical Reports Server (NTRS)

    Kerslake, Thomas W.; Mason, Lee S.; Strumpf, Hal J.

    1997-01-01

    As part of the International Space Station (ISS) Phase 1 program, NASA Lewis Research Center (LERC) and the Russian Space Agency (RSA) teamed together to design, build and flight test the world's first orbital Solar Dynamic Power System (SDPS) on the Russian space station Mir. The Solar Dynamic Flight Demonstration (SDFD) program was to operate a nominal 2 kWe SDPS on Mir for a period up to 1-year starting in late 1997. Unfortunately, the SDFD mission was demanifested from the ISS phase 1 shuttle program in early 1996. However, substantial flight hardware and prototypical flight hardware was built including a heat receiver and aperture shield. The aperture shield comprises the front face of the cylindrical cavity heat receiver and is located at the focal plane of the solar concentrator. It is constructed of a stainless steel plate with a 1-m outside diameter, a 0.24-m inside diameter and covered with high-temperature, refractory metal Multi-Foil Insulation (MFI). The aperture shield must minimize heat loss from the receiver cavity, provide a stiff, high strength structure to accommodate shuttle launch loads and protect receiver structures from highly concentrated solar fluxes during concentrator off-pointing events. To satisfy Mir operational safety protocols, the aperture shield was required to accommodate direct impingement of the intensely concentrated solar image for a 1-hour period. To verify thermal-structural durability under the anticipated high-flux, high-temperature loading, an aperture shield test article was constructed and underwent a series of two tests in a large thermal vacuum chamber configured with a reflective, point-focus solar concentrator and a solar simulator. The test article was positioned near the focal plane and exposed to concentrated solar flux for a period of 1-hour. In the first test, a near equilibrium temperature of 1862 K was attained in the center of the shield hot spot. In the second test, with increased incident flux, a near

  14. Aging Assessment of an Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) Service Cable.

    SciTech Connect

    Bernstein, Robert; Celina, Mathias Christopher; Redline, Erica Marie; White, II, Gregory Von

    2014-07-01

    Nuclear energy is one industry where aging of safety-related materials and components is of great concern. Many U.S. nuclear power plants are approaching, or have already exceeded, 40 years of age. Analysis comparing the cost of new plant construction versus long-term operation under extended plant licensing through 60 years strongly favors the latter option. To ensure the safe, reliable, and cost-effective long-term operation of nuclear power plants, many systems, structures, and components must be evaluated. Furthermore, as new analytical techniques and testing approaches are developed, it is imperative that we also validate, and if necessary, improve upon the previously employed Institute of Electrical and Electronic Engineers (IEEE) qualification standards originally written in 1974. Fortunately, this daunting task has global support, particularly in light of the new social and political climate surrounding nuclear energy in a post-Fukushima era.

  15. Report of the ANS Project Feasibility Workshop for a High Flux Isotope Reactor-Center for Neutron Research Facility

    SciTech Connect

    Peretz, F.J.; Booth, R.S.

    1995-07-01

    The Advanced Neutron Source (ANS) Conceptual Design Report (CDR) and its subsequent updates provided definitive design, cost, and schedule estimates for the entire ANS Project. A recent update to this estimate of the total project cost for this facility was $2.9 billion, as specified in the FY 1996 Congressional data sheet, reflecting a line-item start in FY 1995. In December 1994, ANS management decided to prepare a significantly lower-cost option for a research facility based on ANS which could be considered during FY 1997 budget deliberations if DOE or Congressional planners wished. A cost reduction for ANS of about $1 billion was desired for this new option. It was decided that such a cost reduction could be achieved only by a significant reduction in the ANS research scope and by maximum, cost-effective use of existing High Flux Isotope Reactor (HFIR) and ORNL facilities to minimize the need for new buildings. However, two central missions of the ANS -- neutron scattering research and isotope production-were to be retained. The title selected for this new option was High Flux Isotope Reactor-Center for Neutron Research (HFIR-CNR) because of the project`s maximum use of existing HFIR facilities and retention of selected, central ANS missions. Assuming this shared-facility requirement would necessitate construction work near HFIR, it was specified that HFIR-CNR construction should not disrupt normal operation of HFIR. Additional objectives of the study were that it be highly credible and that any material that might be needed for US Department of Energy (DOE) and Congressional deliberations be produced quickly using minimum project resources. This requirement made it necessary to rely heavily on the ANS design, cost, and schedule baselines. A workshop methodology was selected because assessment of each cost and/or scope-reduction idea required nearly continuous communication among project personnel to ensure that all ramifications of propsed changes.

  16. High Stability Engine Control (HISTEC) Flight Test Results

    NASA Technical Reports Server (NTRS)

    Southwick, Robert D.; Gallops, George W.; Kerr, Laura J.; Kielb, Robert P.; Welsh, Mark G.; DeLaat, John C.; Orme, John S.

    1998-01-01

    The High Stability Engine Control (HISTEC) Program, managed and funded by the NASA Lewis Research Center, is a cooperative effort between NASA and Pratt & Whitney (P&W). The program objective is to develop and flight demonstrate an advanced high stability integrated engine control system that uses real-time, measurement-based estimation of inlet pressure distortion to enhance engine stability. Flight testing was performed using the NASA Advanced Controls Technologies for Integrated Vehicles (ACTIVE) F-15 aircraft at the NASA Dryden Flight Research Center. The flight test configuration, details of the research objectives, and the flight test matrix to achieve those objectives are presented. Flight test results are discussed that show the design approach can accurately estimate distortion and perform real-time control actions for engine accommodation.

  17. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    SciTech Connect

    Daily, Charles R.

    2015-10-01

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  18. Broad-Application Test Reactor

    SciTech Connect

    Motloch, C.G.

    1992-05-01

    This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators.

  19. Test results of the highly instrumented Space Shuttle Main Engine

    NASA Technical Reports Server (NTRS)

    Mcconnaughey, H. V.; Leopard, J. L.; Lightfoot, R. M.

    1992-01-01

    Test results of a highly instrumented Space Shuttle Main Engine (SSME) are presented. The instrumented engine, when combined with instrumented high pressure turbopumps, contains over 750 special measurements, including flowrates, pressures, temperatures, and strains. To date, two different test series, accounting for a total of sixteen tests and 1,667 seconds, have been conducted with this engine. The first series, which utilized instrumented turbopumps, characterized the internal operating environment of the SSME for a variety of operating conditions. The second series provided system-level validation of a high pressure liquid oxygen turbopump that had been retrofitted with a fluid-film bearing in place of the usual pump-end ball bearings. Major findings from these two test series are highlighted in this paper. In addition, comparisons are made between model predictions and measured test data.

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

    SciTech Connect

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

    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.

  1. Advanced Test Reactor Tour

    SciTech Connect

    Miley, Don

    2011-01-01

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

  2. Advanced Test Reactor Tour

    ScienceCinema

    Miley, Don

    2016-07-12

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

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

    SciTech Connect

    Not Available

    1992-07-01

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

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

    SciTech Connect

    Darmann, Frank; Lombaerde, Robert; Moriconi, Franco; Nelson, Albert

    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 wire

  5. Monitoring Delamination of Thermal Barrier Coatings During Interrupted High-Heat-Flux Laser Testing using Luminescence Imaging

    NASA Technical Reports Server (NTRS)

    Eldridge, Jeffrey I.; Zhu, Dongming; Wolfe, Douglas E.

    2011-01-01

    This presentation showed progress made in extending luminescence-base delamination monitoring to TBCs exposed to high heat fluxes, which is an environment that much better simulates actual turbine engine conditions. This was done by performing upconversion luminescence imaging during interruptions in laser testing, where a high-power CO2 laser was employed to create the desired heat flux. Upconverison luminescence refers to luminescence where the emission is at a higher energy (shorter wavelength) than the excitation. Since there will be negligible background emission at higher energies than the excitation, this methods produces superb contrast. Delamination contrast is produced because both the excitation and emission wavelengths are reflected at delamination cracks so that substantially higher luminescence intensity is observed in regions containing delamination cracks. Erbium was selected as the dopant for luminescence specifically because it exhibits upconversion luminescence. The high power CO2 10.6 micron wavelength laser facility at NASA GRC was used to produce the heat flux in combination with forced air backside cooling. Testing was performed at a lower (95 W/sq cm) and higher (125 W/sq cm) heat flux as well as furnace cycling at 1163C for comparison. The lower heat flux showed the same general behavior as furnace cycling, a gradual, "spotty" increase in luminescence associated with debond progression; however, a significant difference was a pronounced incubation period followed by acceleration delamination progression. These results indicate that extrapolating behavior from furnace cycling measurements will grossly overestimate remaining life under high heat flux conditions. The higher heat flux results were not only accelerated, but much different in character. Extreme bond coat rumpling occurred, and delamination propagation extended over much larger areas before precipitating macroscopic TBC failure. This indicates that under the higher heat flux (and

  6. Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

    SciTech Connect

    Knight, R.W.; Morin, R.A.

    1999-12-01

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U{sub 3}O{sub 8} powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated.

  7. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2006

    SciTech Connect

    Primm, R. T.; Ellis, R. J.; Gehin, J. C.; Clarno, K. T.; Williams, K. A.; Moses, D. L.

    2006-11-01

    Neutronics and thermal-hydraulics studies show that, for equivalent operating power [85 MW(t)], a low-enriched uranium (LEU) fuel cycle based on uranium-10 wt % molybdenum (U-10Mo) metal foil with radially, “continuously graded” fuel meat thickness results in a 15% reduction in peak thermal flux in the beryllium reflector of the High Flux Isotope Reactor (HFIR) as compared to the current highly enriched uranium (HEU) cycle. The uranium-235 content of the LEU core is almost twice the amount of the HEU core when the length of the fuel cycle is kept the same for both fuels. Because the uranium-238 content of an LEU core is a factor of 4 greater than the uranium-235 content, the LEU HFIR core would weigh 30% more than the HEU core. A minimum U-10Mo foil thickness of 84 μm is required to compensate for power peaking in the LEU core although this value could be increased significantly without much penalty. The maximum U-10Mo foil thickness is 457μm. Annual plutonium production from fueling the HFIR with LEU is predicted to be 2 kg. For dispersion fuels, the operating power for HFIR would be reduced considerably below 85 MW due to thermal considerations and due to the requirement of a 26-d fuel cycle. If an acceptable fuel can be developed, it is estimated that $140 M would be required to implement the conversion of the HFIR site at Oak Ridge National Laboratory from an HEU fuel cycle to an LEU fuel cycle. To complete the conversion by fiscal year 2014 would require that all fuel development and qualification be completed by the end of fiscal year 2009. Technological development areas that could increase the operating power of HFIR are identified as areas for study in the future.

  8. New neutron small-angle diffraction instrument at the Brookhaven High Flux Beam Reactor

    SciTech Connect

    Schneider, D.K.; Schoenborn, B.P.

    1982-01-01

    The new instrument utilizes cold neutrons emerging from a series of straight neutron guides. A multilayered monochromator is used in combination with a short collimator to obtain a monochromatized beam with a wavelength between 4 and 10 A and a wavelength spread of about 10%. The flux at 5 A exceeds 10/sup 6/ ns/sup -1/ cm/sup -2/ in a typical beam of 6-mm diameter at the sample. The spectrometer itself incorporates provisions for computer-controlled positioning of samples and a two-dimensional detector. At a sample-detector distance between 50 and 200 cm the detector can be centered at scattering angles of up to 45/sup 0/. The beam-defining components, the monochromator, the collimator, and various slits, are easily accessible and exchangeable for alternative devices. These features make the instrument modular and give it flexibility approaching that of standard x-ray equipment.

  9. Test plan for high flux back washable filter at 105 KE fuel storage basin

    SciTech Connect

    Bergsman, K.H.

    1996-08-06

    This document describes the operational testing at the KE Basins planned for demonstrating and evaluating a high flux back washable filter. The document includes the description of the operational testing to be conducted, the samples and measurements to be taken, the sample analysis to be performed, and the use of the information obtained.

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

    SciTech Connect

    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.

  11. Impact of the High Flux Isotope Reactor HEU to LEU Fuel Conversion on Cold Source Nuclear Heat Generation Rates

    SciTech Connect

    Chandler, David

    2014-03-01

    Under the sponsorship of the US Department of Energy National Nuclear Security Administration, staff members at the Oak Ridge National Laboratory have been conducting studies to determine whether the High Flux Isotope Reactor (HFIR) can be converted from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. As part of these ongoing studies, an assessment of the impact that the HEU to LEU fuel conversion has on the nuclear heat generation rates in regions of the HFIR cold source system and its moderator vessel was performed and is documented in this report. Silicon production rates in the cold source aluminum regions and few-group neutron fluxes in the cold source moderator were also estimated. Neutronics calculations were performed with the Monte Carlo N-Particle code to determine the nuclear heat generation rates in regions of the HFIR cold source and its vessel for the HEU core operating at a full reactor power (FP) of 85 MW(t) and the reference LEU core operating at an FP of 100 MW(t). Calculations were performed with beginning-of-cycle (BOC) and end-of-cycle (EOC) conditions to bound typical irradiation conditions. Average specific BOC heat generation rates of 12.76 and 12.92 W/g, respectively, were calculated for the hemispherical region of the cold source liquid hydrogen (LH2) for the HEU and LEU cores, and EOC heat generation rates of 13.25 and 12.86 W/g, respectively, were calculated for the HEU and LEU cores. Thus, the greatest heat generation rates were calculated for the EOC HEU core, and it is concluded that the conversion from HEU to LEU fuel and the resulting increase of FP from 85 MW to 100 MW will not impact the ability of the heat removal equipment to remove the heat deposited in the cold source system. Silicon production rates in the cold source aluminum regions are estimated to be about 12.0% greater at BOC and 2.7% greater at EOC for the LEU core in comparison to the HEU core. Silicon is aluminum s major transmutation product and

  12. Monitoring Delamination of Thermal Barrier Coating During Interrupted High-Heat Flux Laser Testing Using Upconversion Luminescence Imaging

    NASA Technical Reports Server (NTRS)

    Eldridge, Jeffrey I.; Zhu, Dongming; Wolfe, Douglas E.

    2011-01-01

    Upconversion luminescence imaging of thermal barrier coatings (TBCs) has been shown to successfully monitor TBC delamination progression during interrupted furnace cycling. However, furnace cycling does not adequately model engine conditions where TBC-coated components are subjected to significant heat fluxes that produce through-thickness temperature gradients that may alter both the rate and path of delamination progression. Therefore, new measurements are presented based on luminescence imaging of TBC-coated specimens subjected to interrupted high-heat-flux laser cycling exposures that much better simulate the thermal gradients present in engine conditions. The TBCs tested were deposited by electron-beam physical vapor deposition (EB-PVD) and were composed of 7wt% yttria-stabilized zirconia (7YSZ) with an integrated delamination sensing layer composed of 7YSZ co-doped with erbium and ytterbium (7YSZ:Er,Yb). The high-heat-flux exposures that produce the desired through-thickness thermal gradients were performed using a high power CO2 laser operating at a wavelength of 10.6 microns. Upconversion luminescence images revealed the debond progression produced by the cyclic high-heat-flux exposures and these results were compared to that observed for furnace cycling.

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

    SciTech Connect

    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.

  14. Recent High Heat Flux Tests on W-Rod-Armored Mockups

    SciTech Connect

    NYGREN,RICHARD E.; YOUCHISON,DENNIS L.; MCDONALD,JIMMIE M.; LUTZ,THOMAS J.; MISZKIEL,MARK E.

    2000-07-18

    In the authors initial high heat flux tests on small mockups armored with W rods, done in the small electron beam facility (EBTS) at Sandia National Laboratories, the mockups exhibited excellent thermal performance. However, to reach high heat fluxes, they reduced the heated area to only a portion ({approximately}25%) of the sample. They have now begun tests in their larger electron beam facility, EB 1200, where the available power (1.2 MW) is more than enough to heat the entire surface area of the small mockups. The initial results indicate that, at a given power, the surface temperatures of rods in the EB 1200 tests is somewhat higher than was observed in the EBTS tests. Also, it appears that one mockup (PW-10) has higher surface temperatures than other mockups with similar height (10mm) W rods, and that the previously reported values of absorbed heat flux on this mockup were too high. In the tests in EB 1200 of a second mockup, PW-4, absorbed heat fluxes of {approximately}22MW/m{sup 2} were reached but the corresponding surface temperatures were somewhat higher than in EBTS. A further conclusion is that the simple 1-D model initially used in evaluating some of the results from the EBTS testing was not adequate, and 3-D thermal modeling will be needed to interpret the results.

  15. MTR BUILDING INTERIOR, TRA603, REACTOR FLOOR. DETAIL OF REACTOR TEST ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    MTR BUILDING INTERIOR, TRA-603, REACTOR FLOOR. DETAIL OF REACTOR TEST HOLE OPENING IN WEST FACE. CAMERA FACING NORTHEAST. INL NEGATIVE NO. HD46-2-1. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  16. Hypersonic Engine Leading Edge Experiments in a High Heat Flux, Supersonic Flow Environment

    NASA Technical Reports Server (NTRS)

    Gladden, Herbert J.; Melis, Matthew E.

    1994-01-01

    A major concern in advancing the state-of-the-art technologies for hypersonic vehicles is the development of an aeropropulsion system capable of withstanding the sustained high thermal loads expected during hypersonic flight. Three aerothermal load related concerns are the boundary layer transition from laminar to turbulent flow, articulating panel seals in high temperature environments, and strut (or cowl) leading edges with shock-on-shock interactions. A multidisciplinary approach is required to address these technical concerns. A hydrogen/oxygen rocket engine heat source has been developed at the NASA Lewis Research Center as one element in a series of facilities at national laboratories designed to experimentally evaluate the heat transfer and structural response of the strut (or cowl) leading edge. A recent experimental program conducted in this facility is discussed and related to cooling technology capability. The specific objective of the experiment discussed is to evaluate the erosion and oxidation characteristics of a coating on a cowl leading edge (or strut leading edge) in a supersonic, high heat flux environment. Heat transfer analyses of a similar leading edge concept cooled with gaseous hydrogen is included to demonstrate the complexity of the problem resulting from plastic deformation of the structures. Macro-photographic data from a coated leading edge model show progressive degradation over several thermal cycles at aerothermal conditions representative of high Mach number flight.

  17. High heat flux testing of divertor plasma facing materials and components using the HHF test facility at IPR

    NASA Astrophysics Data System (ADS)

    Patil, Yashashri; Khirwadkar, S. S.; Belsare, Sunil; Swamy, Rajamannar; Tripathi, Sudhir; Bhope, Kedar; Kanpara, Shailesh

    2016-02-01

    The High Heat Flux Test Facility (HHFTF) was designed and established recently at Institute for Plasma Research (IPR) in India for testing heat removal capability and operational life time of plasma facing materials and components of the ITER-like tokamak. The HHFTF is equipped with various diagnostics such as IR cameras and IR-pyrometers for surface temperature measurements, coolant water calorimetry for absorbed power measurements and thermocouples for bulk temperature measurements. The HHFTF is capable of simulating steady state heat load of several MW m-2 as well as short transient heat loads of MJ m-2. This paper presents the current status of the HHFTF at IPR and high heat flux tests performed on the curved tungsten monoblock type of test mock-ups as well as transient heat flux tests carried out on pure tungsten materials using the HHFTF. Curved tungsten monoblock type of test mock-ups were fabricated using hot radial pressing (HRP) technique. Two curved tungsten monoblock type test mock-ups successfully sustained absorbed heat flux up to 14 MW m-2 with thermal cycles of 30 s ON and 30 s OFF duration. Transient high heat flux tests or thermal shock tests were carried out on pure tungsten hot-rolled plate material (Make:PLANSEE) with incident power density of 0.49 GW m-2 for 20 milliseconds ON and 1000 milliseconds OFF time. A total of 6000 thermal shock cycles were completed on pure tungsten material. Experimental results were compared with mathematical simulations carried out using COMSOL Multiphysics for transient high heat flux tests.

  18. Preliminary considerations of an intense slow positron facility based on a sup 78 Kr loop in the high flux isotopes reactor

    SciTech Connect

    Hulett, L.D. Jr.; Donohue, D.L.; Peretz, F.J.; Montgomery, B.H.; Hayter, J.B.

    1990-01-01

    Suggestions have been made to the National Steering Committee for the Advanced Neutron Source (ANS) by Mills that provisions be made to install a high intensity slow positron facility, based on a {sup 78}Kr loop, that would be available to the general community of scientists interested in this field. The flux of thermal neutrons calculated for the ANS is E + 15 sec{sup {minus}1} m{sup {minus}2}, which Mills has estimated will produce 5 mm beam of slow positrons having a current of about 1 E + 12 sec {sup {minus}1}. The intensity of such a beam will be a least 3 orders of magnitude greater than those presently available. The construction of the ANS is not anticipated to be complete until the year 2000. In order to properly plan the design of the ANS, strong considerations are being given to a proof-of-principle experiment, using the presently available High Flux Isotopes Reactor, to test the {sup 78}Kr loop technique. The positron current from the HFIR facility is expected to be about 1 E + 10 sec{sup {minus}1}, which is 2 orders of magnitude greater than any other available. If the experiment succeeds, a very valuable facility will be established, and important formation will be generated on how the ANS should be designed. 3 refs., 1 fig.

  19. Study of the Potential Impact of Gamma-Induced Radiolytic Gases on Loading of Cesium Onto Crystalline Silicotitanate Sorbent at ORNL's High Flux Isotope Reactor

    SciTech Connect

    Mattus, A.J.

    2001-02-12

    The use of an engineered form of crystalline silicotitanate as a potential sorbent for the removal and concentration of cesium from the high-level waste at the Savannah River Site was investigated. Results conclusively showed this sorbent to be unaffected by gamma-induced radiolytic gas formation during column loading. Closely controlled column-loading experiments were performed at the Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR) in a gamma field with a conservative dose rate expected to exceed that in a full-scale column by a factor of nearly 16. Operation of column loading under expected nominal full-scale field conditions in the HFIR pool showed that radiolytic gases were formed at a previously calculated generation rate of 0.4 mL per liter of feed solution. When the resulting cesium-loading curve in the gamma field was compared with that of a control experiment in the absence of a gamma field, no discernable difference in the curves (within analytical error) was detected. Both curves were in good agreement with the VERSE computer-generated curve. Results conclusively indicate that the production of radiolytic gases within a full-scale column is not expected to result in reduced capacity or associated gas generation problems during operation at the Savannah River Site.

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

    SciTech Connect

    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.

  1. Results of high heat flux tests of tungsten divertor targets under plasma heat loads expected in ITER and tokamaks (review)

    NASA Astrophysics Data System (ADS)

    Budaev, V. P.

    2016-12-01

    Heat loads on the tungsten divertor targets in the ITER and the tokamak power reactors reach 10MW m-2 in the steady state of DT discharges, increasing to 0.6-3.5 GW m-2 under disruptions and ELMs. The results of high heat flux tests (HHFTs) of tungsten under such transient plasma heat loads are reviewed in the paper. The main attention is paid to description of the surface microstructure, recrystallization, and the morphology of the cracks on the target. Effects of melting, cracking of tungsten, drop erosion of the surface, and formation of corrugated and porous layers are observed. Production of submicron-sized tungsten dust and the effects of the inhomogeneous surface of tungsten on the plasma-wall interaction are discussed. In conclusion, the necessity of further HHFTs and investigations of the durability of tungsten under high pulsed plasma loads on the ITER divertor plates, including disruptions and ELMs, is stressed.

  2. Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

    SciTech Connect

    Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N.

    2006-02-01

    A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

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

    SciTech Connect

    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).

  4. Space Shuttle Main Engine instrumented High Pressure Oxidizer Turbopump technology test bed testing results summary

    NASA Technical Reports Server (NTRS)

    Koelbl, Mary E.

    1993-01-01

    This paper presents the test results from the Space Shuttle Main Engine (SSME) instrumented High Pressure Oxidizer Turbopump (HPOTP). The turbopump was tested on Engine 3001, a highly instrumented engine, in an effort to characterize the turbopump and the engine system. Seven tests, for a total duration of 766 seconds, were performed over a five month time period. The testing was performed at a wide variety of engine conditions. Changes in engine mixture ratio, power level, engine inlet oxidizer pressure, engine inlet fuel pressure, and engine start sequence were made. A discussion of all the HPOTP pressure and temperature data obtained are presented with comparisons to supporting analyses made where applicable. The effect of the various engine conditions on the measured data is addressed. This paper also discusses the challenges that were overcome to obtain the data. The significant instrumentation related problems encountered during the design, fabrication, and testing of this turbopump are summarized. Only those issues that affected the data obtained or the instrumentation itself are discussed. The relevance of the data to other noninstrumented turbomachinery is outlined. Conclusions and recommendations resulting from the test series will be presented.

  5. Final Report Independent Verification Survey of the High Flux Beam Reactor, Building 802 Fan House Brookhaven National Laboratory Upton, New York

    SciTech Connect

    Harpeneau, Evan M.

    2011-06-24

    On May 9, 2011, ORISE conducted verification survey activities including scans, sampling, and the collection of smears of the remaining soils and off-gas pipe associated with the 802 Fan House within the HFBR (High Flux Beam Reactor) Complex at BNL. ORISE is of the opinion, based on independent scan and sample results obtained during verification activities at the HFBR 802 Fan House, that the FSS (final status survey) unit meets the applicable site cleanup objectives established for as left radiological conditions.

  6. Examination of loop-operator-initiated events for the advanced test reactor

    SciTech Connect

    Durney, J.L.; Majumdar, D.

    1989-01-01

    The Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory is a unique high-flux test reactor having nine major test positions for irradiation of reactor materials. These test positions contain inpile tubes (IPT) that are connected to external piping and equipment (loops) to provide the high-temperature, high-pressure environment for the testing. The design of the core has intimately integrated the IPTs into the fuel region by means of a serpentine fuel arrangement resulting in a close reactivity coupling between the loop thermal hydraulics and the core. Consequently, operator actions potentially have an impact on the reactor power transients resulting from off-normal conditions in these facilities. This paper examines these operator-initiated events and their consequences. The analysis of loop-operator-initiated events indicates there is no damage to the reactor core even when assuming no operator intervention for mitigation. However, analysis does assume a scram occurs when required by the reactor protection systems.

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

    SciTech Connect

    Yoder Jr, Graydon L; Aaron, Adam M; Cunningham, Richard Burns; Fugate, David L; Holcomb, David Eugene; Kisner, Roger A; Peretz, Fred J; Robb, Kevin R; Wilgen, John B; Wilson, Dane F

    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.

  8. Tests of the Daimler D-IVa Engine at a High Altitude Test Bench

    NASA Technical Reports Server (NTRS)

    Noack, W G

    1920-01-01

    Reports of tests of a Daimler IVa engine at the test-bench at Friedrichshafen, show that the decrease of power of that engine, at high altitudes, was established, and that the manner of its working when air is supplied at a certain pressure was explained. These tests were preparatory to the installation of compressors in giant aircraft for the purpose of maintaining constant power at high altitudes.

  9. Spheromak reactor with poloidal flux-amplifying transformer

    DOEpatents

    Furth, Harold P.; Janos, Alan C.; Uyama, Tadao; Yamada, Masaaki

    1987-01-01

    An inductive transformer in the form of a solenoidal coils aligned along the major axis of a flux core induces poloidal flux along the flux core's axis. The current in the solenoidal coil is then reversed resulting in a poloidal flux swing and the conversion of a portion of the poloidal flux to a toroidal flux in generating a spheromak plasma wherein equilibrium approaches a force-free, minimum Taylor state during plasma formation, independent of the initial conditions or details of the formation. The spheromak plasma is sustained with the Taylor state maintained by oscillating the currents in the poloidal and toroidal field coils within the plasma-forming flux core. The poloidal flux transformer may be used either as an amplifier stage in a moving plasma reactor scenario for initial production of a spheromak plasma or as a method for sustaining a stationary plasma and further heating it. The solenoidal coil embodiment of the poloidal flux transformer can alternately be used in combination with a center conductive cylinder aligned along the length and outside of the solenoidal coil. This poloidal flux-amplifying inductive transformer approach allows for a relaxation of demanding current carrying requirements on the spheromak reactor's flux core, reduces plasma contamination arising from high voltage electrode discharge, and improves the efficiency of poloidal flux injection.

  10. Production of medical radioisotopes in the ORNL High Flux Isotope Reactor (HFIR) for cancer treatment and arterial restenosis therapy after PTCA

    SciTech Connect

    Knapp, F.F. Jr.; Beets, A.L.; Mirzadeh, S.; Alexander, C.W.; Hobbs, R.L.

    1998-06-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) represents an important resource for the production of a wide variety of medical radioisotopes. In addition to serving as a key production site for californium-252 and other transuranic elements, important examples of therapeutic radioisotopes which are currently routinely produced in the HFIR for distribution include dysprosium-166 (parent of holmium-166), rhenium-186, tin-117m and tungsten-188 (parent of rhenium-188). The nine hydraulic tube (HT) positions in the central high flux region permit the insertion and removal of targets at any time during the operating cycle and have traditionally represented a major site for production of medical radioisotopes. To increase the irradiation capabilities of the HFIR, special target holders have recently been designed and fabricated which will be installed in the six Peripheral Target Positions (PTP), which are also located in the high flux region. These positions are only accessible during reactor refueling and will be used for long-term irradiations, such as required for the production of tin-117m and tungsten-188. Each of the PTP tubes will be capable of housing a maximum of eight HT targets, thus increasing the total maximum number of HT targets from the current nine, to a total of 57. In this paper the therapeutic use of reactor-produced radioisotopes for bone pain palliation and vascular brachytherapy and the therapeutic medical radioisotope production capabilities of the ORNL HFIR are briefly discussed.

  11. Production of Medical Radioisotopes in the ORNL High Flux Isotope Reactor (HFIR) for Cancer Treatment and Arterial Restenosis Therapy after PTCA

    DOE R&D Accomplishments Database

    Knapp, F. F. Jr.; Beets, A. L.; Mirzadeh, S.; Alexander, C. W.; Hobbs, R. L.

    1998-06-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) represents an important resource for the production of a wide variety of medical radioisotopes. In addition to serving as a key production site for californium-252 and other transuranic elements, important examples of therapeutic radioisotopes which are currently routinely produced in the HFIR for distribution include dysprosium-166 (parent of holmium-166), rhenium-186, tin-117m and tungsten-188 (parent of rhenium-188). The nine hydraulic tube (HT) positions in the central high flux region permit the insertion and removal of targets at any time during the operating cycle and have traditionally represented a major site for production of medical radioisotopes. To increase the irradiation capabilities of the HFIR, special target holders have recently been designed and fabricated which will be installed in the six Peripheral Target Positions (PTP), which are also located in the high flux region. These positions are only accessible during reactor refueling and will be used for long-term irradiations, such as required for the production of tin-117m and tungsten-188. Each of the PTP tubes will be capable of housing a maximum of eight HT targets, thus increasing the total maximum number of HT targets from the current nine, to a total of 57. In this paper the therapeutic use of reactor-produced radioisotopes for bone pain palliation and vascular brachytherapy and the therapeutic medical radioisotope production capabilities of the ORNL HFIR are briefly discussed.

  12. On0Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactor

    SciTech Connect

    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.

  13. Cu-Cr-Nb-Zr Alloy for Rocket Engines and Other High-Heat- Flux Applications

    NASA Technical Reports Server (NTRS)

    Ellis, David L.

    2013-01-01

    Rocket-engine main combustion chamber liners are used to contain the burning of fuel and oxidizer and provide a stream of high-velocity gas for propulsion. The liners in engines such as the Space Shuttle Main Engine are regeneratively cooled by flowing fuel, e.g., cryogenic hydrogen, through cooling channels in the back side of the liner. The heat gained by the liner from the flame and compression of the gas in the throat section is transferred to the fuel by the liner. As a result, the liner must either have a very high thermal conductivity or a very high operating temperature. In addition to the large heat flux (>10 MW/sq m), the liners experience a very large thermal gradient, typically more than 500 C over 1 mm. The gradient produces thermally induced stresses and strains that cause low cycle fatigue (LCF). Typically, a liner will experience a strain differential in excess of 1% between the cooling channel and the hot wall. Each time the engine is fired, the liner undergoes an LCF cycle. The number of cycles can be as few as one for an expendable booster engine, to as many as several thousand for a reusable launch vehicle or reaction control system. Finally, the liners undergo creep and a form of mechanical degradation called thermal ratcheting that results in the bowing out of the cooling channel into the combustion chamber, and eventual failure of the liner. GRCop-84, a Cu-Cr-Nb alloy, is generally recognized as the best liner material available at the time of this reporting. The alloy consists of 14% Cr2Nb precipitates in a pure copper matrix. Through experimental work, it has been established that the Zr will not participate in the formation of Laves phase precipitates with Cr and Nb, but will instead react with Cu to form the desired Cu-Zr compounds. It is believed that significant improvements in the mechanical properties of GRCop-84 will be realized by adding Zr. The innovation is a Cu-Cr-Nb-Zr alloy covering the composition range of 0.8 to 8.1 weight

  14. Thermal hydraulic analysis of China fusion engineering test reactor during thermal quenching by comparative approach of Relap5 and THEATRe codes

    NASA Astrophysics Data System (ADS)

    Khan, Salah Ud-Din; Song, Yuntao; Khan, Shahab Ud-Din

    2016-10-01

    Thermal quenching in Tokamak reactor is the most obvious phenomenon happens during plasma disruption conditions. The current research is focused on the thermal behavior of different parameters of China fusion engineering test reactor (CFETR) including reactor power, pressure and mass flow rate conditions. The analysis was performed by two thermal hydraulic codes, i.e. THEATRe and Relap5. During the first phase of research, thermal quenching behavior and trends that can be possible during the reactor operation was performed. In the next phase, nodalization diagram of THEATRe and Relap5 codes were developed. The listed parameters were calculated and analyzed for the safety aspects of the reactor. The main objective of the research was to analyze the blanket system of CFETR (Tokamak) for safety concerns during disruption condition. The research will be extended to other components for safe operation of reactor as well.

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

    SciTech Connect

    Yoder Jr, Graydon L; Elkassabgi, Yousri M.; De Leon, Gerardo I.; Fetterly, Caitlin N.; Ramos, Jorge A.; Cunningham, Richard Burns

    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

  16. Energy Efficient Engine high pressure turbine component test performance report

    NASA Technical Reports Server (NTRS)

    Timko, L. P.

    1984-01-01

    The high pressure turbine for the General Electric Energy Efficient Engine is a two stage design of moderate loading. Results of detailed system studies led to selection of this configuration as the most appropriate in meeting the efficiency goals of the component development program. To verify the design features of the high pressure turbine, a full scale warm air turbine test rig with cooling flows simulated was run. Prior to this testing, an annular cascade test was run to select vane unguided turn for the first stage nozzle. Results of this test showed that the base configuration exceeded the lower unguided turning configuration by 0.48 percent in vane kinetic energy efficiency. The air turbine test program, consisting of extensive mapping and cooling flow variation as well as design point evaluation, demonstrated a design point efficiency level of 90.0 percent based on the thermodynamic definition. In terms of General Electric cycle definition, this efficiency was 92.5 percent. Based on this test, it is concluded that efficiency goals for the Flight Propulsion System were met.

  17. Fast Flux Test Facility final safety analysis report. Amendment 73

    SciTech Connect

    Gantt, D.A.

    1993-08-01

    This report provides Final Safety Analysis Report (FSAR) Amendment 73 for incorporation into the Fast Flux Test Facility (FFTR) FSAR set. This page change incorporates Engineering Change Notices (ECNs) issued subsequent to Amendment 72 and approved for incorparoration before May 6, 1993. These changes include: Chapter 3, design criteria structures, equipment, and systems; chapter 5B, reactor coolant system; chapter 7, instrumentation and control systems; chapter 9, auxiliary systems; chapter 11, reactor refueling system; chapter 12, radiation protection and waste management; chapter 13, conduct of operations; chapter 17, technical specifications; chapter 20, FFTF criticality specifications; appendix C, local fuel failure events; and appendix Fl, operation at 680{degrees}F inlet temperature.

  18. Thermal modelling of various thermal barrier coatings in a high heat flux rocket engine

    NASA Technical Reports Server (NTRS)

    Nesbitt, James A.

    1989-01-01

    Traditional Air Plasma Sprayed (APS) ZrO2-Y2O3 Thermal Barrier Coatings (TBC's) and Low Pressure Plasma Sprayed (LPPS) ZrO2-Y2O3/Ni-Cr-Al-Y cermet coatings were tested in a H2/O2 rocked engine. The traditional ZrO2-Y2O3 (TBC's) showed considerable metal temperature reductions during testing in the hydrogen-rich environment. A thermal model was developed to predict the thermal response of the tubes with the various coatings. Good agreement was observed between predicted temperatures and measured temperatures at the inner wall of the tube and in the metal near the coating/metal interface. The thermal model was also used to examine the effect of the differences in the reported values of the thermal conductivity of plasma sprayed ZrO2-Y2O3 ceramic coatings, the effect of 100 micron (0.004 in.) thick metallic bond coat, the effect of tangential heat transfer around the tube, and the effect or radiation from the surface of the ceramic coating. It was shown that for the short duration testing in the rocket engine, the most important of these considerations was the effect of the uncertainty in the thermal conductivity of temperatures (greater than 100 C) predicted in the tube. The thermal model was also used to predict the thermal response of the coated rod in order to quantify the difference in the metal temperatures between the two substrate geometries and to explain the previously-observed increased life of coatings on rods over that on tubes. A thermal model was also developed to predict heat transfer to the leading edge of High Pressure Fuel Turbopump (HPFTP) blades during start-up of the space shuttle main engines. The ability of various TBC's to reduce metal temperatures during the two thermal excursions occurring on start-up was predicted. Temperature reductions of 150 to 470 C were predicted for 165 micron (0.0065 in.) coatings for the greater of the two thermal excursions.

  19. Advanced Test Reactor -- Testing Capabilities and Plans AND Advanced Test Reactor National Scientific User Facility -- Partnerships and Networks

    SciTech Connect

    Frances M. Marshall

    2008-07-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world’s premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner “lobes” to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. For future research, some ATR modifications and enhancements are currently planned. In 2007 the US Department of Energy designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR for material testing research by a broader user community. This paper provides more details on some of the ATR capabilities, key design features, experiments, and plans for the NSUF.

  20. ETRCF, TRA654, INTERIOR. TEST VESSEL (NOT REACTOR) INSIDE PIT. INL ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    ETR-CF, TRA-654, INTERIOR. TEST VESSEL (NOT REACTOR) INSIDE PIT. INL NEGATIVE NO. HD24-2-2. Mike Crane, Photographer, ca. 2003 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  1. High heat flux performance of brazed tungsten macro-brush test mock-up for divertors

    NASA Astrophysics Data System (ADS)

    Patil, Yashashri; Khirwadkar, S. S.; Krishnan, D.; Patel, A.; Tripathi, S.; Singh, K. P.; Belsare, S. M.

    2013-06-01

    Plasma facing components (PFCs) of divertor will be exposed to steady state and transient heat loads up to 20 MW/m2, during operation of ITER-like plasma fusion device. The critical task in fusion research is to design, fabricate and test of PFCs. To withstand high heat loads, PFCs are designed and fabricated in flat tile, mono-block type geometries using tungsten as plasma facing material and CuCrZr alloy is used as a heat sink. These fabricated mock-ups are tested under thermal cyclic heat loads using intense electron beam in pulsed mode. Tungsten macro-brush type of mock-up has been developed by vacuum furnace brazing route. Mock-up was tested to the absorbed heat flux in the range of 0.5-9 MW/m2. Simulation of high heat flux (HHF) test under steady state and cyclic heat loads has been done using ANSYS12 finite element analysis (FEA) software. HHF tests have been successfully performed on the tungsten mock-up.

  2. Research and Development of Multiphysics Models in Support of the Conversion of the High Flux Isotope Reactor to Low Enriched Uranium Fuel

    SciTech Connect

    Bodey, Isaac T.; Curtis, Franklin G.; Arimilli, Rao V.; Ekici, Kivanc; Freels, James D.

    2015-11-01

    The findings presented in this report are results of a five year effort led by the RRD Division of the ORNL, which is focused on research and development toward the conversion of the High Flux Isotope Reactor (HFIR) fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU). This report focuses on the tasks accomplished by the University of Tennessee Knoxville (UTK) team from the Department of Mechanical, Aerospace, and Biomedical Engineering (MABE) that provided expert support in multiphysics modeling of complex problems associated with the LEU conversion of the HFIR reactor. The COMSOL software was used as the main computational modeling tool, whereas Solidworks was also used in support of computer-aided-design (CAD) modeling of the proposed LEU fuel design. The UTK research has been governed by a statement of work (SOW), which was updated annually to clearly define the specific tasks reported herein. Ph.D. student Isaac T. Bodey has focused on heat transfer and fluid flow modeling issues and has been aided by his major professor Dr. Rao V. Arimilli. Ph.D. student Franklin G. Curtis has been focusing on modeling the fluid-structure interaction (FSI) phenomena caused by the mechanical forces acting on the fuel plates, which in turn affect the fluid flow in between the fuel plates, and ultimately the heat transfer, is also affected by the FSI changes. Franklin Curtis has been aided by his major professor Dr. Kivanc Ekici. M.Sc. student Adam R. Travis has focused two major areas of research: (1) on accurate CAD modeling of the proposed LEU plate design, and (2) reduction of the model complexity and dimensionality through interdimensional coupling of the fluid flow and heat transfer for the HFIR plate geometry. Adam Travis is also aided by his major professor, Dr. Kivanc Ekici. We must note that the UTK team, and particularly the graduate students, have been in very close collaboration with Dr. James D. Freels (ORNL technical monitor and mentor) and have

  3. LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor: Appendix A-4

    SciTech Connect

    Sviatoslavsky, I.N.; Attaya, H.M.; Corradini, M.L.; Lomperski, S.

    1987-01-01

    This paper describes the preliminary analysis of LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor breeding shield. TIBER-II is a compact reactor with a major radius of 3 m and thus requires a thin, high efficiency shield on the inboard side. The use of tungsten in the inboard shield implies a rather high rate of afterheat upon plasma shutdown, which must be dissipated in a controlled manner to avoid the possibility of radioactivity release or threatening the investment. Because the shield is cooled with an aqueous solution, LOFA does not pose a problem as long as natural convection can be established. LOCA, however, has more serious consequences, particularly on the inboard side. Circulation of air by natural convection is proposed as a means for dissipating the inboard shield decay heat. The safety and environmental implications of such a scheme are evaluated. It is shown that the inboard shield temperature never exceeds 510/sup 0/C following LOCA posing no hazard to reactor personnel and not threatening the investment. 7 refs., 6 figs.

  4. A high flux pulsed source of energetic atomic oxygen. [for spacecraft materials ground testing

    NASA Technical Reports Server (NTRS)

    Krech, Robert H.; Caledonia, George E.

    1986-01-01

    The design and demonstration of a pulsed high flux source of nearly monoenergetic atomic oxygen are reported. In the present test setup, molecular oxygen under several atmospheres of pressure is introduced into an evacuated supersonic expansion nozzle through a pulsed molecular beam valve. A 10J CO2 TEA laser is focused to intensities greater than 10 to the 9th W/sq cm in the nozzle throat, generating a laser-induced breakdown with a resulting 20,000-K plasma. Plasma expansion is confined by the nozzle geometry to promote rapid electron-ion recombination. Average O-atom beam velocities from 5-13 km/s at fluxes up to 10 to the 18th atoms/pulse are measured, and a similar surface oxygen enrichment in polyethylene samples to that obtained on the STS-8 mission is found.

  5. Type B investigation of the iridium contamination event at the High Flux Isotope Reactor on September 7, 1993

    SciTech Connect

    Not Available

    1994-03-01

    On the title date, at ORNL, area radiation alarms sounded during a routine transfer of a shielding cask (containing 60 Ci{sup 192}Ir) from the HFIR pool side to a transport truck. Small amounts of Ir were released from the cask onto the reactor bay floor. The floor was cleaned, and the cask was shipped to a hot cell at Building 3047 on Oct. 3, 1993. The event was caused by rupture of one of the Ir target rods after it was loaded into the cask for normal transport operations; the rupture was the result of steam generation in the target rod soon after it was placed in the cask (water had entered the target rod through a tiny defect in a weld while it was in the reactor under pressure). While the target rods were in the reactor and reactor pool, there was sufficient cooling to prevent steam generation; when the target rod was loaded into the dry transport cask, the temperature increased enough to result in boiling of the trapped water and produced high enough pressure to result in rupture. The escaping steam ejected some of the Ir pellets. The event was reported as Occurrence Report Number ORO--MMES-X10HFIR-1993-0030, dated Sept. 8, 1993. Analysis indicated that the following conditions were probable causes: less than adequate welding procedures, practices, or techniques, material controls, or inspection methods, or combination thereof, could have led to weld defects, affecting the integrity of target rod IR-75; less than adequate secondary containment in the cask allowed Ir pellets to escape.

  6. High temperature reactors

    NASA Astrophysics Data System (ADS)

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

    2008-12-01

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

  7. Use and Storage of Test and Operations Data from the High Temperature Test Reactor Acquired by the US Government from the Japan Atomic Energy Agency

    SciTech Connect

    Hans Gougar

    2010-02-01

    This document describes the use and storage of data from the High Temperature Test Reactor (HTTR) acquired from the Japan Atomic Energy Agency (JAEA) by the U.S. Government for high temperature reactor research under the Next Generation Nuclear Plant (NGNP) Project.

  8. Current and future capabilities of the neutron reflectometer MIRROR at Oak Ridge National Laboratory's High Flux Isotope Reactor

    NASA Astrophysics Data System (ADS)

    Hamilton, W. A.; Smith, G. S.; Taylor, G. B.; Larkins, B. M.; Porcar, L.

    2006-11-01

    The peripatetic ORNL HFIR Center for Neutron Scattering reflectometer instrument MIRROR has recently been re-installed in an interim beam line position in the reactor beam room. In 2006 an upgraded version of the instrument will move to a high intensity guide hall position fed by the new HFIR cold source. In this short note, we present some aspects of current instrument operation-particularly with respect to data reduction from the instrument's linear reflection plane detector-with examples of ongoing research and analysis, and a brief outline of the expected capabilities of the fully upgraded guide hall instrument.

  9. Use of high-radiant flux, high-resolution DMD light engines in industrial applications

    NASA Astrophysics Data System (ADS)

    Müller, Alexandra; Ram, Surinder

    2014-03-01

    The field of application of industrial projectors is growing day by day. New Digital Micromirror Device (DMD) - based applications like 3D printing, 3D scanning, Printed Circuit Board (PCB) board printing and others are getting more and more sophisticated. The technical demands for the projection system are rising as new and more stringent requirements appear. The specification for industrial projection systems differ substantially from the ones of business and home beamers. Beamers are designed to please the human eye. Bright colors and image enhancement are far more important than uniformity of the illumination or image distortion. The human eye, followed by the processing of the brain can live with quite high intensity variations on the screen and image distortion. On the other hand, a projector designed for use in a specialized field has to be tailored regarding its unique requirements in order to make no quality compromises. For instance, when the image is projected onto a light sensitive resin, a good uniformity of the illumination is crucial for good material hardening (curing) results. The demands on the hardware and software are often very challenging. In the following we will review some parameters that have to be considered carefully for the design of industrial projectors in order to get the optimum result without compromises.

  10. Effects of Test Cell Recirculation on High-Bypass Turbofan Engines during Simulated Altitude Tests

    DTIC Science & Technology

    1986-08-01

    methods used to describe and correct for test cell recirculation have led to this investigation, which is in line with the objectives previously stated...between pressure profiles on the cell wall and those on the model surfaces, which is in line with previous observations. Removal of the diffuser...AEDC-TR-85-55 , , . ’ . . . .L Effects of Test Cell Recirculation on High-Bypass Turbofan Engines During Simulated Altitude Tests OCT 0 9

  11. Internal Combustion Engines as Fluidized Bed Reactors

    NASA Astrophysics Data System (ADS)

    Lavich, Zoe; Taie, Zachary; Menon, Shyam; Beckwith, Walter; Daly, Shane; Halliday, Devin; Hagen, Christopher

    2016-11-01

    Using an internal combustion engine as a chemical reactor could provide high throughput, high chemical conversion efficiency, and reactant/product handling benefits. For processes requiring a solid catalyst, the ability to develop a fluidized bed within the engine cylinder would allow efficient processing of large volumes of fluid. This work examines the fluidization behavior of particles in a cylinder of an internal combustion engine at various engine speeds. For 40 micron silica gel particles in a modified Megatech Mark III transparent combustion engine, calculations indicate that a maximum engine speed of about 60.8 RPM would result in fluidization. At higher speeds, the fluidization behavior is expected to deteriorate. Experiments gave qualitative confirmation of the analytical predictions, as a speed of 48 RPM resulted in fluidized behavior, while a speed of 171 RPM did not. The investigation shows that under certain conditions a fluidized bed can be obtained within an engine cylinder. Corresponding Author.

  12. Neutronics Simulations of 237Np Targets to Support Safety-Basis and 238Pu Production Assessment Efforts at the High Flux Isotope Reactor

    SciTech Connect

    Chandler, David; Ellis, Ronald James

    2015-01-01

    Fueled by two highly enriched uranium-bearing fuel elements surrounded by a large concentric ring of beryllium reflector, the High Flux Isotope Reactor (HFIR) provides one of the highest neutron fluxes in the world and is used to produce unique isotopes like plutonium-238. The National Aeronautics and Space Administration use radioisotope thermoelectric generators powered by 238Pu for deep-space missions. As part of the US Department of Energy s task to reestablish the domestic production of 238Pu, a technology demonstration sub-project has been initiated to establish a new 238Pu supply chain. HFIR safety-basis neutronics calculations are being performed to ensure the target irradiations have no adverse impacts on reactor performance and to calculate data required as input to follow-on thermal-structural, thermal-hydraulic and radionuclide/dose analyses. Plutonium-238 production assessments are being performed to estimate the amount of 238Pu that can be produced in HFIR s permanent beryllium reflector. It is estimated that a total of 0.96 1.12 kg 238Pu (~1.28 1.49 kg PuO2 at 85% 238Pu/Pu purity) could be produced per year in HFIR s permanent beryllium reflector irradiation facilities if they are all utilized.

  13. Type A verification report for the high flux beam reactor stack and grounds, Brookhaven National Laboratory, Upton, New York

    SciTech Connect

    Harpenau, Evan M.

    2012-01-13

    The U.S. Department of Energy (DOE) Order 458.1 requires independent verification (IV) of DOE cleanup projects (DOE 2011). The Oak Ridge Institute for Science and Education (ORISE) has been designated as the responsible organization for IV of the High Flux Beam Reactor (HFBR) Stack and Grounds area at Brookhaven National Laboratory (BNL) in Upton, New York. The IV evaluation may consist of an in-process inspection with document and data reviews (Type A Verification) or a confirmatory survey of the site (Type B Verification). DOE and ORISE determined that a Type A verification of the documents and data for the HFBR Stack and Grounds: Survey Units (SU) 6, 7, and 8 was appropriate based on the initial survey unit classification, the walkover surveys, and the final analytical results provided by the Brookhaven Science Associates (BSA). The HFBR Stack and Grounds surveys began in June 2011 and were completed in September 2011. Survey activities by BSA included gamma walkover scans and sampling of the as-left soils in accordance with the BSA Work Procedure (BNL 2010a). The Field Sampling Plan - Stack and Remaining HFBR Outside Areas (FSP) stated that gamma walk-over surveys would be conducted with a bare sodium iodide (NaI) detector, and a collimated detector would be used to check areas with elevated count rates to locate the source of the high readings (BNL 2010b). BSA used the Mult- Agency Radiation Survey and Site Investigation Manual (MARSSIM) principles for determining the classifications of each survey unit. Therefore, SUs 6 and 7 were identified as Class 1 and SU 8 was deemed Class 2 (BNL 2010b). Gamma walkover surveys of SUs 6, 7, and 8 were completed using a 2X2 NaI detector coupled to a data-logger with a global positioning system (GPS). The 100% scan surveys conducted prior to the final status survey (FSS) sampling identified two general soil areas and two isolated soil locations with elevated radioactivity. The general areas of elevated activity identified

  14. High altitude aerodynamic platform concept evaluation and prototype engine testing

    NASA Technical Reports Server (NTRS)

    Akkerman, J. W.

    1984-01-01

    A design concept has been developed for maintaining a 150-pound payload at 60,000 feet altitude for about 50 hours. A 600-pound liftoff weight aerodynamic vehicle is used which operates at sufficient speeds to withstand prevailing winds. It is powered by a turbocharged four-stoke cycle gasoline fueled engine. Endurance time of 100 hours or more appears to be feasible with hydrogen fuel and a lighter payload. A prototype engine has been tested to 40,000 feet simulated altitude. Mismatch of the engine and the turbocharger system flow and problems with fuel/air mixture ratio control characteristics prohibited operation beyond 40,000 feet. But there seems to be no reason why the concept cannot be developed to function as analytically predicted.

  15. Advanced Test Reactor National Scientific User Facility Partnerships

    SciTech Connect

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

    2012-03-01

    -Madison; (8) Illinois Institute of Technology (IIT) Materials Research Collaborative Access Team (MRCAT) beamline at Argonne National Laboratory's Advanced Photon Source; and (9) Nanoindenter in the University of California at Berkeley (UCB) Nuclear Engineering laboratory Materials have been analyzed for ATR NSUF users at the Advanced Photon Source at the MRCAT beam, the NIST Center for Neutron Research in Gaithersburg, MD, the Los Alamos Neutron Science Center, and the SHaRE user facility at Oak Ridge National Laboratory (ORNL). Additionally, ORNL has been accepted as a partner facility to enable ATR NSUF users to access the facilities at the High Flux Isotope Reactor and related facilities.

  16. Operational safety at the fast flux test facility

    SciTech Connect

    Bennett, C.L.; Baird, Q.L.; Franz, G.R.

    1986-01-01

    The safety organization within Westinghouse Hanford Company (WHC) provides the independent review and appraisal of reactor facilities at the Hanford Engineering Development Laboratory (HEDL) in accordance with US Department of Energy (DOE) Order 5480.1A, Chapter V. The safety organization functions primarily in an advisory capacity to the line organization and reports through a management organization independent of all reactor operations to the president of WHC. However, safety is a line responsibility, and neither review nor subsequent approval by the safety staff releases line management from its responsibility for the safety of people and equipment. The purpose of this paper is to describe the operational safety program at HEDL associated with the operation of the Fast Flux Test Facility (FFTF). These activities include: (1) operational reactor safety surveillance; (2) change review of safety documentation; (3) cycle readiness assessments; (4) FFTF technical specification upgrade; (5) interim examination and maintenance cell and fuel storage facility safety review.

  17. High-density reduced-enrichment fuels for Research and Test Reactors

    SciTech Connect

    Snelgrove, J.L.; Hofman, G.L.; Copeland, G.L.

    1983-01-01

    Development and irradiation testing of high-density fuels have been conducted by the US RERTR Program in order to provide the technical means to reduce the enrichment of fuels for research and test reactors. The traditional aluminum dispersion fuel technology has been extended to include the highest practical loadings of uranium-aluminide (UAl/sub x/, 2.3 MgU/m/sup 3/), uranium-oxide (U/sub 3/O/sub 8/, 3.2 MgU/m/sup 3/), and uranium-silicide (U/sub 3/Si/sub 2/, 5.5 MgU/m/sup 3/; U/sub 3/Si, 7.0 MgU/m/sup 3/) fuels. A third uranium-silicide alloy, U/sub 3/SiAl (U + 3.5 wt % Si + 1.5 wt % Al) has been found to perform poorly at high burnup. Testing of miniature fuel plates and full-sized fuel elements is at an advanced stage for the highest loadings of the aluminide and oxide fuels and intermediate loadings of the silicide fuels, and good results have been obtained for low-enriched uranium. The data obtained to date are discussed. 1 reference, 3 figures, 1 table.

  18. High frequency data acquisition system for space shuttle main engine testing

    NASA Technical Reports Server (NTRS)

    Lewallen, Pat

    1987-01-01

    The high frequency data acquisition system developed for the Space Shuttle Main Engine (SSME) single engine test facility at the National Space Technology Laboratories is discussed. The real time system will provide engineering data for a complete set of SSME instrumentation (approx. 100 measurements) within 4 hours following engine cutoff, a decrease of over 48 hours from the previous analog tape based system.

  19. Reactor Simulator Testing

    NASA Technical Reports Server (NTRS)

    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.

  20. Thermal response of various thermal barrier coatings in a high heat flux rocket engine

    NASA Technical Reports Server (NTRS)

    Nesbitt, J. A.

    1990-01-01

    Traditional APS ZrO2-Y2O3 thermal barrier coatings (TBCs) formed by air plasma spraying and low pressure and air plasma sprayed ZrO2-Y2O3/NiCrAlY cermet coatings were tested in an H2-O2 rocket engine. The test cycle was approximately 1.2 s at 1400 C in a hydrogen-rich environment. During testing, the maximum metal temperature without a coating was 1310 C. The traditional ZrO2-Y2O3 TBCs with a 100-125 micron thick ceramic layer reduced the maximum metal temperature by approximately 350 C. Increasing the ceramic layer thickness to 200-225 microns resulted in an additional metal temperature reduction of 100 C. However, the cermet coatings, consisting of a ceramic and metal mixture, exhibited a much lower thermal protection capability by reducing the maximum metal temperature by approximately 100 C. It was also found that the surface roughness of the traditional TBCs had little effect on the thermal response.

  1. On flux effects in a low alloy steel from a Swedish reactor pressure vessel

    NASA Astrophysics Data System (ADS)

    Boåsen, Magnus; Efsing, Pål; Ehrnstén, Ulla

    2017-02-01

    This study aims to investigate the presence of Unstable Matrix Defects in irradiated pressure vessel steel from weldments of the Swedish PWR Ringhals 4 (R4). Hardness tests have been performed on low flux (surveillance material) and high flux (Halden reactor) irradiated material samples in combination with heat treatments at temperatures of 330, 360 and 390 °C in order to reveal eventual recovery of any hardening features induced by irradiation. The experiments carried out in this study could not reveal any hardness recovery related to Unstable Matrix Defects at relevant temperatures. However, a difference in hardness recovery was found between the low and the high flux samples at heat treatments at higher temperatures than expected for the annihilation of Unstable Matrix Defects-the observed recovery is here attributed to differences of the solute clusters formed by the high and low flux irradiations.

  2. Markets for reactor-produced non-fission radioisotopes

    SciTech Connect

    Bennett, R.G.

    1995-01-01

    Current market segments for reactor produced radioisotopes are developed and reported from a review of current literature. Specific radioisotopes studied in is report are the primarily selected from those with major medical or industrial markets, or those expected to have strongly emerging markets. Relative market sizes are indicated. Special emphasis is given to those radioisotopes that are best matched to production in high flux reactors such as the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory or the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. A general bibliography of medical and industrial radioisotope applications, trends, and historical notes is included.

  3. Deterministic Modeling of the High Temperature Test Reactor with DRAGON-HEXPEDITE

    SciTech Connect

    J. Ortensi; M.A. Pope; R.M. Ferrer; J.J. Cogliati; J.D. Bess; A.M. Ougouag

    2010-10-01

    The 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 the INL’s current prismatic reactor 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 fuel column thin annular core, and the fully loaded core critical condition with 30 fuel columns. Special emphasis is devoted to physical phenomena and artifacts in HTTR that are similar to phenomena and artifacts in the NGNP base design. The DRAGON code is used in this study since it offers significant ease and versatility in modeling prismatic designs. 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. The results from this study show reasonable agreement in the calculation of the core multiplication factor with the MC methods, but a consistent bias of 2–3% with the experimental values is obtained. 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 partially stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral

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

    SciTech Connect

    Fredrickson, E. D.; Bell, M. G.; Budny, R. V.; Darrow, D. S.; White, R.

    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.

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

    NASA Astrophysics Data System (ADS)

    Fredrickson, E. D.; Bell, M. G.; Budny, R. V.; Darrow, D. S.; White, R.

    2015-03-01

    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 βn ≈ 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.

  6. High-flux focusable color-tunable and efficient white-light-emitting diode light engine for stage lighting

    NASA Astrophysics Data System (ADS)

    Chakrabarti, Maumita; Pedersen, Henrik Chresten; Petersen, Paul Michael; Poulsen, Christian; Poulsen, Peter Behrensdorff; Dam-Hansen, Carsten

    2016-08-01

    A color mixing light-emitting diode (LED) light engine that can replace 2-kW halogen-Fresnel spotlight with high-luminous flux in excess of 20,000 lm is reported for applications in professional stage and studio lighting. The light engine focuses and mixes the light from 210 LEDs of five different colors through a microlens array (MA) at the gate of Ø50 mm. Hence, it produces homogeneous color-mixed tunable white light from 3000 to 6000 K that can be adjustable from flood to spot position providing 10% translational loss, whereas the corresponding loss from the halogen-Fresnel spotlight is 37%. The design, simulation, and optimization of the light engine is described and compared to the experimental characterization of a prototype. The light engine is optimized through the simulated design of reflector, total internal reflection lens, and MA, as well as the number of LEDs. An optical efficiency of 59% and a luminous efficacy of 33 lm/W are achieved, which is three times higher than the 2-kW halogen-Fresnel spotlight. In addition to having color rendering of color rendering index Ra>85 and television lighting consistency index 12>70, the dimmable and tunable white light can be color controlled during the operational time.

  7. MATERIALS TESTING REACTOR (MTR) BUILDING, TRA603. CONTEXTUAL VIEW OF MTR ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    MATERIALS TESTING REACTOR (MTR) BUILDING, TRA-603. CONTEXTUAL VIEW OF MTR BUILDING SHOWING NORTH SIDES OF THE HIGH-BAY REACTOR BUILDING, ITS SECOND/THIRD FLOOR BALCONY LEVEL, AND THE ATTACHED ONE-STORY OFFICE/LABORATORY BUILDING, TRA-604. CAMERA FACING SOUTHEAST. VERTICAL CONCRETE-SHROUDED BEAMS SUPPORT PRECAST CONCRETE PANELS. CONCRETE PROJECTION FORMED AS A BUNKER AT LEFT OF VIEW IS TRA-657, PLUG STORAGE BUILDING. INL NEGATIVE NO. HD46-42-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  8. Thermal Modelling of Various Thermal Barrier Coatings in a High Flux Rocket Engine

    NASA Technical Reports Server (NTRS)

    Nesbitt, James A.

    1998-01-01

    A thermal model was developed to predict the thermal response of coated and uncoated tubes tested in a H2/O2 rocket engine. Temperatures were predicted for traditional APS ZrO2-Y2O3 thermal barrier coatings, as well as APS and LPPS ZrO2-Y2O3/NiCrAlY cermet coatings. Good agreement was observed between predicted and measured metal temperatures at locations near the tube surface or at the inner tube wall. The thermal model was also used to quantitatively examine the effect of various coating system parameters on the temperatures in the substrate and coating. Accordingly, the effect of the presence a metallic bond coat and the effect of radiation from the surface of the ceramic layer were examined. In addition, the effect of a variation in the values of the thermal conductivity of the ceramic layer was also investigated. It was shown that a variation in the thermal conductivity of the ceramic layer, on the order of that reported in the literature for plasma sprayed ZrO2-Y2O3 coatings, can result in temperature differences in the substrate greater than 100 C, a much greater effect than that due to the presence of a bond coat or radiation from the ceramic layer. The thermal model was also used to predict the thermal response of a coated rod in order to quantify the difference in the metal temperatures between the two substrate geometries in order to explain the previously-observed increased life of coatings on rods over that on tubes. It was shown that for the short duration testing in the rocket engine, the temperature in a tube could exceed that in a rod by more than 100 C. Lastly, a two-dimensional model was developed to evaluate the effect of tangential heat transfer around the tube and its impact on reducing the stagnation point temperature. It was also shown that tangential heat transfer does not significantly reduce the stagnation point temperature, thus allowing application of a simpler, one-dimensional model for comparing measured and predicted stagnation point

  9. TYPE A VERIFICATION REPORT FOR THE HIGH FLUX BEAM REACTOR STACK AND GROUNDS, BROOKHAVEN NATIONAL LABORATORY, UPTON, NEW YORK DCN 5098-SR-08-0

    SciTech Connect

    Evan Harpenau

    2011-11-30

    The U.S. Department of Energy (DOE) Order 458.1 requires independent verification (IV) of DOE cleanup projects (DOE 2011). The Oak Ridge Institute for Science and Education (ORISE) has been designated as the responsible organization for IV of the High Flux Beam Reactor (HFBR) Stack and Grounds area at Brookhaven National Laboratory (BNL) in Upton, New York. The IV evaluation may consist of an in-process inspection with document and data reviews (Type A Verification) or a confirmatory survey of the site (Type B Verification). DOE and ORISE determined that a Type A verification of the documents and data for the HFBR Stack and Grounds: Survey Units (SU) 6, 7, and 8 was appropriate based on the initial survey unit classification, the walkover surveys, and the final analytical results provided by the Brookhaven Science Associates (BSA).

  10. Reactor Simulator Testing Overview

    NASA Technical Reports Server (NTRS)

    Schoenfeld, Michael P.

    2013-01-01

    Test Objectives Summary: a) Verify operation of the core simulator, the instrumentation & control system, and the ground support gas and vacuum test equipment. b) Examine cooling & heat regeneration performance of the cold trap purification. c) Test the ALIP pump at voltages beyond 120V to see if the targeted mass flow rate of 1.75 kg/s can be obtained in the RxSim. Testing Highlights: a) Gas and vacuum ground support test equipment performed effectively for operations (NaK fill, loop pressurization, and NaK drain). b) Instrumentation & Control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings and ramped within prescribed constraints. It effectively interacted with reactor simulator control model and defaulted back to temperature control mode if the transient fluctuations didn't dampen. c) 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 minimum temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

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

    SciTech Connect

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

    2015-07-01

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

  12. Reactor Simulator Testing

    NASA Technical Reports Server (NTRS)

    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.

  13. RELAP5/MOD2. 5 analysis of the HFBR (High Flux Beam Reactor) for a loss of power and coolant accident

    SciTech Connect

    Slovik, G.C.; Rohatgi, U.S.; Jo, Jae.

    1990-05-01

    A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs.

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

    NASA Astrophysics Data System (ADS)

    Flanders, T. Michael; Sparks, Mary Helen; Daniel, Joshua D.

    2016-02-01

    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.

  15. Long Distance Reactor Antineutrino Flux Monitoring

    NASA Astrophysics Data System (ADS)

    Dazeley, Steven; Bergevin, Marc; Bernstein, Adam

    2015-10-01

    The feasibility of antineutrino detection as an unambiguous and unshieldable way to detect the presence of distant nuclear reactors has been studied. While KamLAND provided a proof of concept for long distance antineutrino detection, the feasibility of detecting single reactors at distances greater than 100 km has not yet been established. Even larger detectors than KamLAND would be required for such a project. Considerations such as light attenuation, environmental impact and cost, which favor water as a detection medium, become more important as detectors get larger. We have studied both the sensitivity of water based detection media as a monitoring tool, and the scientific impact such detectors might provide. A next generation water based detector may be able to contribute to important questions in neutrino physics, such as supernova neutrinos, sterile neutrino oscillations, and non standard electroweak interactions (using a nearby compact accelerator source), while also providing a highly sensitive, and inherently unshieldable reactor monitoring tool to the non proliferation community. In this talk I will present the predicted performance of an experimental non proliferation and high-energy physics program. Lawrence Livermore National Laboratory is operated by Lawrence Livermore National Security, LLC, for the U.S. Department of Energy, National Nuclear Security Administration under Contract DE-AC52-07NA27344. Release number LLNL-ABS-674192.

  16. Assessment of current structural design methodology for high-temperature reactors based on failure tests

    SciTech Connect

    Corum, J.M.; Sartory, W.K.

    1985-01-01

    A mature design methodology, consisting of inelastic analysis methods, provided in Department of Energy guidelines, and failure criteria, contained in ASME Code Case N-47, exists in the United States for high-temperature reactor components. The objective of this paper is to assess the adequacy of this overall methodology by comparing predicted inelastic deformations and lifetimes with observed results from structural failure tests and from an actual service failure. Comparisons are presented for three types of structural situations: (1) nozzle-to-spherical shell specimens, where stresses at structural discontinuities lead to cracking, (2) welded structures, where metallurgical discontinuities play a key role in failures, and (3) thermal shock loadings of cylinders and pipes, where thermal discontinuities can lead to failure. The comparison between predicted and measured inelastic responses are generally reasonalbly good; quantities are sometimes overpredicted somewhat, and, sometimes underpredicted. However, even seemingly small discrepancies can have a significant effect on structural life, and lifetimes are not always as closely predicted. For a few cases, the lifetimes are substantially overpredicted, which raises questions regarding the adequacy of existing design margins.

  17. Radiological transportation risk assessment of the shipment of sodium-bonded fuel from the Fast Flux Test Facility to the Idaho National Engineering Laboratory

    SciTech Connect

    Green, J.R.

    1995-01-31

    This document was written in support of Environmental Assessment: Shutdown of the Fast Flux Test Facility (FFTF), Hanford Site, Richland, Washington. It analyzes the potential radiological risks associated with the transportation of sodium-bonded metal alloy and mixed carbide fuel from the FFTF on the Hanford Site in Washington State to the Idaho Engineering Laboratory in Idaho in the T-3 Cask. RADTRAN 4 is used for the analysis which addresses potential risk from normal transportation and hypothetical accident scenarios.

  18. High-flux years

    SciTech Connect

    Not Available

    1992-01-01

    This article reviews the history of activities at the Clinton Laboratories from the end of World War II into roughly mid 1948. This was a period of great change for the laboratory. Monsanto Chemical Company who had served as the industrial operator of the laboratory was replaced by an academic team from the University of Chicago, which never really got started, and was replaced by a new industrial operator, Union Carbide Corporation. Clinton Laboratories became Clinton National Laboratory in 1947 and Oak Ridge National Laboratory in 1948. In the face of these management uncertainties and upheavals much productive work and cutting science was achieved. The Materials Testing Reactor was designed, and served as the model for light-water reactors. The Graphite Reactor began routine production of radioisotopes. New groups were formed to begin and strengthen work in biology, metallurgy, and health physics.

  19. High heat flux test with HIP-bonded Ferritic Martensitic Steel mock-up for the first wall of the KO HCML TBM

    NASA Astrophysics Data System (ADS)

    Won Lee, Dong; Dug Bae, Young; Kwon Kim, Suk; Yun Shin, Hee; Guen Hong, Bong; Cheol Bang, In

    2011-10-01

    In order for a Korean Helium Cooled Molten Lithium (HCML) Test Blanket Module (TBM) to be tested in the International Thermonuclear Experimental Reactor (ITER), fabrication method for the TBM FW such as Hot Isostatic Pressing (HIP, 1050 °C, 100 MPa, 2 h) has been developed including post HIP heat treatment (PHHT, normalizing at 950 °C for 2 h and tempering at 750 °C for 2 h) with Ferritic Martensitic Steel (FMS). Several mock-ups were fabricated using the developed methods and one of them, three-channel mock-up, was used for performing a High Heat Flux (HHF) test to verify the joint integrity. Test conditions were determined using the commercial code, ANSYS-11, and the test was performed in the Korea Heat Load Test (KoHLT) facility, which was used a radiation heating with a graphite heater. The mock-up survived up to 1000 cycles under 1.0 MW/m 2 heat flux and there is no delamination or failure during the test.

  20. Solvent extraction studies with high-burnup Fast Flux Test Facility fuel in the Solvent Extraction Test Facility

    SciTech Connect

    Benker, D.E.; Bigelow, J.E.; Bond, W.D.; Chattin, F.R.; King, L.J.; Kitts, F.G.; Ross, R.G.; Stacy, R.G.

    1986-10-01

    A batch of high-burnup fuel from the Fast Flux Test Facility (FFTF) was processed in the Solvent Extraction Test Facility (SETF) during Campaign 9. The fuel had a burnup of {similar_to}0 MWd/kg and a cooling time of {similar_to} year. Two runs were made with this fuel; in the first, the solvent contained 30% tri-n-butyl phosphate (TBP) and partitioning of the uranium and plutonium was effected by reducing the plutonium with hydroxylamine nitrate (HAN); in the second, the solvent contained 10% TBP and a low operating temperature was used in an attempt to partition without reducing the plutonium valence. The plutonium reoxidation problem, which was present in previous runs that used HAN, may have been solved by lowering the temperature and acidity in the partition contactor. An automatic control system was used to maintain high loadings of heavy metals in the coextraction-coscrub contactor in order to increase its efficiency while maintaining low losses of uranium and plutonium to the aqueous raffinate. An in-line photometer system was used to measure the plutonium concentration in an intermediate extraction stage; and based on this data, a computer algorithm determined the appropriate adjustments in the addition rate of the extractant. The control system was successfully demonstrated in a preliminary run with purified uranium. However, a variety of equipment and start up problems prevented an extended demonstration from being accomplished during the runs with the FFTF fuel.

  1. Temperature controlled material irradiation in the advanced test reactor

    NASA Astrophysics Data System (ADS)

    Ingram, F. W.; Palmer, A. J.; Stites, D. J.

    1998-10-01

    The United States Department of Energy (US DOE) has initiated the development of an Irradiation Test Vehicle (ITV) for fusion materials irradiation at the Advanced Test Reactor (ATR) in Idaho Falls, Idaho, USA. The ITV is capable of providing neutron spectral tailoring and individual temperature control for up to 15 experiment capsules simultaneously. The test vehicle consists of three In-Pile Tubes (IPTs) running the length of the reactor vessel. These IPTs are kept dry and test trains with integral instrumentation are inserted and removed through a transfer shield plate above the reactor vessel head. The test vehicle is designed to irradiate specimens as large as 2.2 cm in diameter, at temperatures of 250-800°C, achieving neutron damage rates as high as 10 displacements per atom per year. The high fast to thermal neutron flux ratio required for fusion materials testing is accomplished by using an aluminum filler to displace as much water as possible from the flux trap and surrounding the filler piece with a ring of replaceable neutron absorbing material. The gas blend temperature control system remains in place from test to test, thus hardware costs for new tests are limited to the experiment capsule train and integral instrumentation.

  2. Fibre-optic gamma-flux monitoring in a fission reactor by means of Cerenkov radiation

    NASA Astrophysics Data System (ADS)

    Brichard, B.; Fernandez, A. F.; Ooms, H.; Berghmans, F.

    2007-10-01

    We demonstrate the possibility of using Cerenkov radiation to monitor the reactor power and the high energy gamma-ray flux in a high neutron flux reactor. The system employs a radiation-resistant pure silica glass fibre to measure the Cerenkov radiation in the infrared region (800-1100 nm). A model is proposed to determine the order of magnitude of the gamma-ray flux from the measurement. The method and concept can be extended to the monitoring of low reactor powers if Cerenkov radiation is measured in the 450-500 nm region by means of hydrogen-treated fibres.

  3. Design of a Gas Test Loop Facility for the Advanced Test Reactor

    SciTech Connect

    C. A. Wemple

    2005-09-01

    The Office of Nuclear Energy within the U.S. Department of Energy (DOE-NE) has identified the need for irradiation testing of nuclear fuels and materials, primarily in support of the Generation IV (Gen-IV) and Advanced Fuel Cycle Initiative (AFCI) programs. These fuel development programs require a unique environment to test and qualify potential reactor fuel forms. This environment should combine a high fast neutron flux with a hard neutron spectrum and high irradiation temperature. An effort is presently underway at the Idaho National Laboratory (INL) to modify a large flux trap in the Advanced Test Reactor (ATR) to accommodate such a test facility [1,2]. The Gas Test Loop (GTL) Project Conceptual Design was initiated to determine basic feasibility of designing, constructing, and installing in a host irradiation facility, an experimental vehicle that can replicate with reasonable fidelity the fast-flux test environment needed for fuels and materials irradiation testing for advanced reactor concepts. Such a capability will be needed if programs such as the AFCI, Gen-IV, the Next Generation Nuclear Plant (NGNP), and space nuclear propulsion are to meet development objectives and schedules. These programs are beginning some irradiations now, but many call for fast flux testing within this decade.

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

    SciTech Connect

    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.

  5. High Altitude Small Engine Test Techniques at the NASA Glenn Propulsion Systems Lab

    NASA Technical Reports Server (NTRS)

    Castner, Raymond; Wyzykowski, John; Chiapetta, Santo

    2001-01-01

    A High Altitude Test was performed in the Propulsion Systems Lab (PSL) at the NASA Glenn Research Center using a Pratt and Whitney Canada PW545 jet engine. This engine was tested to develop a highaltitude database on small, high-bypass ratio, engine performance and operability. Industry is interested in the use of high-bypass engines for Uninhabited Aerial Vehicles (UAV's) to perform high altitude surveillance. The tests were a combined effort between Pratt & Whitney Canada (PWC) and NASA Glenn Research Center. A large portion of this test activity was to collect performance data with a highly instrumented low-pressure turbine. Low-pressure turbine aerodynamic performance at low Reynolds numbers was collected and compared to analytical models developed by NASA and PWC. This report describes the test techniques implemented to obtain high accuracy turbine performance data in an altitude test facility, including high accuracy airflow at high altitudes, very low mass flow, and low air temperatures. Major accomplishments from this test activity were to collect accurate and repeatable turbine performance data at high altitudes to within 1 percent. Data were collected at 19,800m, 16,750m, and 13,700m providing documentation of diminishing LPT performance with reductions in Reynolds number in an actual engine flight environment. The test provided a unique database for the development of engine analysis codes to be used for future LPT performance improvements.

  6. FFTF (Fast Flux Test Facility) as an irradiation test bed for fusion materials and components

    SciTech Connect

    Greenslade, D.L.; Puigh, R.J.; Hollenberg, G.W.; Grover, J.M.

    1986-03-01

    The relatively large irradiation volume, instrumentation capabilities, and fast neutron flux associated with the Fast Flux Test Facility (FFTF) make this reactor an ideal test bed for fusion materials and components irradiations. Significant fusion materials irradiations are presently being performed in the Materials Open Test Assembly (MOTA) in FFTF. The MOTA is providing a controlled temperature and high neutron flux environment for such materials as the low activation alloys, copper alloys, ceramic insulators, and high heat flux materials. Conceptual designs utilizing the versatile MOTA irradiation vehicle have been developed to investigate irradiation effects on the mechanical and tritium breeding behaviors of solid breeder materials. More aggressive conceptual designs have also been developed to irradiate solid breeder blanket submodules in the FFTF. These specific component test designs will be presented and their potential roles in the development of fusion technology discussed.

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

    SciTech Connect

    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

  8. Spring 2007 MCAS High School Science and Technology/Engineering Tests: Summary of State Results

    ERIC Educational Resources Information Center

    Massachusetts Department of Education, 2007

    2007-01-01

    In spring 2007, four Massachusetts Comprehensive Assessment System (MCAS) Science and Technology/Engineering (STE) operational tests were introduced at the high school level (grades 9 and 10): Biology, Chemistry, Introductory Physics, and Technology/Engineering. Over 100,000 Massachusetts public high school students in grades 9 and 10 participated…

  9. FAST FLUX TEST FACILITY (FFTF) A HISTORY OF SAFETY & OPERATIONAL EXCELLENCE

    SciTech Connect

    NIELSEN, D L

    2004-02-26

    The Fast Flux Test Facility (FFTF) is a 400-megawatt (thermal) sodium-cooled, high temperature, fast neutron flux, loop-type test reactor. The facility was constructed to support development and testing of fuels, materials and equipment for the Liquid Metal Fast Breeder Reactor program. FFTF began operation in 1980 and over the next 10 years demonstrated its versatility to perform experiments and missions far beyond the original intent of its designers. The reactor had several distinctive features including its size, flux, core design, extensive instrumentation, and test features that enabled it to simultaneously carry out a significant array of missions while demonstrating its features that contributed to a high level of plant safety and availability. FFTF is currently being deactivated for final closure.

  10. Test and evaluation of the HIDEC engine uptrim algorithm. [Highly Integrated Digital Electronic Control for aircraft

    NASA Technical Reports Server (NTRS)

    Ray, R. J.; Myers, L. P.

    1986-01-01

    The highly integrated digital electronic control (HIDEC) program will demonstrate and evaluate the improvements in performance and mission effectiveness that result from integrated engine-airframe control systems. Performance improvements will result from an adaptive engine stall margin mode, a highly integrated mode that uses the airplane flight conditions and the resulting inlet distortion to continuously compute engine stall margin. When there is excessive stall margin, the engine is uptrimmed for more thrust by increasing engine pressure ratio (EPR). The EPR uptrim logic has been evaluated and implemente into computer simulations. Thrust improvements over 10 percent are predicted for subsonic flight conditions. The EPR uptrim was successfully demonstrated during engine ground tests. Test results verify model predictions at the conditions tested.

  11. High density operation for reactor-relevant power exhaust

    NASA Astrophysics Data System (ADS)

    Wischmeier, M.

    2015-08-01

    With increasing size of a tokamak device and associated fusion power gain an increasing power flux density towards the divertor needs to be handled. A solution for handling this power flux is crucial for a safe and economic operation. Using purely geometric arguments in an ITER-like divertor this power flux can be reduced by approximately a factor 100. Based on a conservative extrapolation of current technology for an integrated engineering approach to remove power deposited on plasma facing components a further reduction of the power flux density via volumetric processes in the plasma by up to a factor of 50 is required. Our current ability to interpret existing power exhaust scenarios using numerical transport codes is analyzed and an operational scenario as a potential solution for ITER like divertors under high density and highly radiating reactor-relevant conditions is presented. Alternative concepts for risk mitigation as well as strategies for moving forward are outlined.

  12. TEST REACTOR AREA PLOT PLAN CA. 1968. MTR AND ETR ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    TEST REACTOR AREA PLOT PLAN CA. 1968. MTR AND ETR AREAS SOUTH OF PERCH AVENUE. "COLD" SERVICES NORTH OF PERCH. ADVANCED TEST REACTOR IN NEW SECTION WEST OF COLD SERVICES SECTION. NEW PERIMETER FENCE ENCLOSES BETA RAY SPECTROMETER, TRA-669, AN ATR SUPPORT FACILITY, AND ATR STACK. UTM LOCATORS HAVE BEEN DELETED. IDAHO NUCLEAR CORPORATION, FROM A BLAW-KNOX DRAWING, 3/1968. INL INDEX NO. 530-0100-00-400-011646, REV. 0. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

    SciTech Connect

    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.

  14. High-heat-flux testing of irradiated tungsten-based materials for fusion applications using infrared plasma arc lamps

    SciTech Connect

    Sabau, Adrian S.; Ohriner, Evan K.; Kiggans, Jim; Schaich, Charles R.; Ueda, Yoshio; Harper, David C.; Katoh, Yutai; Snead, Lance L.; Byun, Thak S.

    2014-11-01

    Testing of advanced materials and component mock-ups under prototypical fusion high-heat-flux conditions, while historically a mainstay of fusion research, has proved to be quite challenging, especially for irradiated materials. A new high-heat-flux–testing (HHFT) facility based on water-wall plasma arc lamps (PALs) is now introduced for materials and small-component testing. Two PAL systems, utilizing a 12 000°C plasma arc contained in a quartz tube cooled by a spiral water flow over the inside tube surface, provide maximum incident heat fluxes of 4.2 and 27 MW/m2 over areas of 9×12 and 1×10 cm2, respectively. This paper will present the overall design and implementation of a PAL-based irradiated material target station (IMTS). The IMTS is primarily designed for testing the effects of heat flux or thermal cycling on material coupons of interest, such as those for plasma-facing components. Temperature results are shown for thermal cycling under HHFT of tungsten coupon specimens that were neutron irradiated in HFIR. Finally, radiological surveys indicated minimal contamination of the 36×36×18 cm test section, demonstrating the capability of the new facility to handle irradiated specimens at high temperature.

  15. High-heat-flux testing of irradiated tungsten-based materials for fusion applications using infrared plasma arc lamps

    DOE PAGES

    Sabau, Adrian S.; Ohriner, Evan K.; Kiggans, Jim; ...

    2014-11-01

    Testing of advanced materials and component mock-ups under prototypical fusion high-heat-flux conditions, while historically a mainstay of fusion research, has proved to be quite challenging, especially for irradiated materials. A new high-heat-flux–testing (HHFT) facility based on water-wall plasma arc lamps (PALs) is now introduced for materials and small-component testing. Two PAL systems, utilizing a 12 000°C plasma arc contained in a quartz tube cooled by a spiral water flow over the inside tube surface, provide maximum incident heat fluxes of 4.2 and 27 MW/m2 over areas of 9×12 and 1×10 cm2, respectively. This paper will present the overall design andmore » implementation of a PAL-based irradiated material target station (IMTS). The IMTS is primarily designed for testing the effects of heat flux or thermal cycling on material coupons of interest, such as those for plasma-facing components. Temperature results are shown for thermal cycling under HHFT of tungsten coupon specimens that were neutron irradiated in HFIR. Finally, radiological surveys indicated minimal contamination of the 36×36×18 cm test section, demonstrating the capability of the new facility to handle irradiated specimens at high temperature.« less

  16. Enhanced In-Pile Instrumentation at the Advanced Test Reactor

    SciTech Connect

    Joy Rempe; Darrell Knudson; Joshua Daw; Troy Unruh; Benjamin Chase; Kurt Davis; Robert Schley; Steven Taylor

    2012-08-01

    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 provides an update on 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 real-time 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.

  17. Enhanced In-Pile Instrumentation at the Advanced Test Reactor

    SciTech Connect

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

    2011-06-01

    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 (INL). To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility (NSUF) 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 real-time 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.

  18. Titer plate formatted continuous flow thermal reactors for high throughput applications: fabrication and testing

    NASA Astrophysics Data System (ADS)

    Sang-Won Park, Daniel; Chen, Pin-Chuan; You, Byoung Hee; Kim, Namwon; Park, Taehyun; Lee, Tae Yoon; Datta, Proyag; Desta, Yohannes; Soper, Steven A.; Nikitopoulos, Dimitris E.; Murphy, Michael C.

    2010-05-01

    A high throughput, multi-well (96) polymerase chain reaction (PCR) platform, based on a continuous flow (CF) mode of operation, was developed. Each CFPCR device was confined to a footprint of 8 × 8 mm2, matching the footprint of a well on a standard micro-titer plate. While several CFPCR devices have been demonstrated, this is the first example of a high-throughput multi-well continuous flow thermal reactor configuration. Verification of the feasibility of the multi-well CFPCR device was carried out at each stage of development from manufacturing to demonstrating sample amplification. The multi-well CFPCR devices were fabricated by micro-replication in polymers, polycarbonate to accommodate the peak temperatures during thermal cycling in this case, using double-sided hot embossing. One side of the substrate contained the thermal reactors and the opposite side was patterned with structures to enhance thermal isolation of the closely packed constant temperature zones. A 99 bp target from a λ-DNA template was successfully amplified in a prototype multi-well CFPCR device with a total reaction time as low as ~5 min at a flow velocity of 3 mm s-1 (15.3 s cycle-1) and a relatively low amplification efficiency compared to a bench-top thermal cycler for a 20-cycle device; reducing the flow velocity to 1 mm s-1 (46.2 s cycle-1) gave a seven-fold improvement in amplification efficiency. Amplification efficiencies increased at all flow velocities for 25-cycle devices with the same configuration.

  19. Engine Tests Using High-Sulfur Diesel Fuel

    DTIC Science & Technology

    1980-09-01

    REPORT DOCUMENTATION PAGE BEFORE COMPLETINQ FORM- 1. REPORT NUMBER ~2.GOVT ACCESSION NO, 3. RECIPIENT’S CATALOG NUMBIER AFLRL No. 129 .. A )A-)0 iij_...SHEETS. 0 Prior to the test, calibrate: tachometer flowmeter load measurement system temperature indicators (1-7) and exhaust temperature (1 and 2

  20. Initial progress in the first wall, blanket, and shield Engineering Test Program for magnetically confined fusion-power reactors

    SciTech Connect

    Herman, H.; Baker, C.C.; Maroni, V.A.

    1981-10-01

    The first wall/blanket/shield (FW/B/S) Engineering Test Program (ETP) progressed from the planning stage into implementation during July, 1981. The program, generic in nature, comprises four Test Program Elements (TPE's), the emphasis of which is on defining the performance parameters for the Fusion Engineering Device (FED) and the major fusion device to follow FED. These elements are: (1) nonnuclear thermal-hydraulic and thermomechanical testing of first wall and component facsimiles with emphasis on surface heat loads and heat transient (i.e., plasma disruption) effects; (2) nonnuclear and nuclear testing of FW/B/S components and assemblies with emphasis on bulk (nuclear) heating effects, integrated FW/B/S hydraulics and mechanics, blanket coolant system transients, and nuclear benchmarks; (3) FW/B/S electromagnetic and eddy current effects testing, including pulsed field penetration, torque and force restraint, electromagnetic materials, liquid metal MHD effects and the like; and (4) FW/B/S Assembly, Maintenance and Repair (AMR) studies focusing on generic AMR criteria, with the objective of preparing an AMR designers guidebook; also, development of rapid remote assembly/disassembly joint system technology, leak detection and remote handling methods.

  1. Hyper-X Engine Testing in the NASA Langley 8-Foot High Temperature Tunnel

    NASA Technical Reports Server (NTRS)

    Huebner, Lawrence D.; Rock, Kenneth E.; Witte, David W.; Ruf, Edward G.; Andrews, Earl H., Jr.

    2000-01-01

    Airframe-integrated scramjet engine tests have 8 completed at Mach 7 in the NASA Langley 8-Foot High Temperature Tunnel under the Hyper-X program. These tests provided critical engine data as well as design and database verification for the Mach 7 flight tests of the Hyper-X research vehicle (X-43), which will provide the first-ever airframe- integrated scramjet flight data. The first model tested was the Hyper-X Engine Model (HXEM), and the second was the Hyper-X Flight Engine (HXFE). The HXEM, a partial-width, full-height engine that is mounted on an airframe structure to simulate the forebody features of the X-43, was tested to provide data linking flowpath development databases to the complete airframe-integrated three-dimensional flight configuration and to isolate effects of ground testing conditions and techniques. The HXFE, an exact geometric representation of the X-43 scramjet engine mounted on an airframe structure that duplicates the entire three-dimensional propulsion flowpath from the vehicle leading edge to the vehicle base, was tested to verify the complete design as it will be flight tested. This paper presents an overview of these two tests, their importance to the Hyper-X program, and the significance of their contribution to scramjet database development.

  2. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-3: High Temperature Gas Cooled Reactor Thermal-Hydraulics.

    ERIC Educational Resources Information Center

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical high temperature gas-cooled reactor (HTGR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module…

  3. Space reactor ground tests, assessment of facility needs

    SciTech Connect

    Philbin, J.S.; Wemple, R.P.

    1985-01-01

    This paper discusses facility requirements for previously-specified, space-reactor ground tests of operating reactor prototypes. The paper also discusses engineering development tests applicable to fully-integrated space reactors and their subsystems. The development tests were derived by considering the functions and environments that the reactor might encounter in normal or abnormal launch-to-recovery sequences. Surety (safety, safeguards, and reliability) priorities will influence the final character of both the test program and the facilities requirements. Sandia facilities suitable for conducting the above tests are the main focus of this paper.

  4. Evaluation of Cooling Conditions for a High Heat Flux Testing Facility Based on Plasma-Arc Lamps

    SciTech Connect

    Charry, Carlos H.; Abdel-khalik, Said I.; Yoda, Minami; Sabau, Adrian S.; Snead, Lance Lewis

    2015-07-31

    The new Irradiated Material Target Station (IMTS) facility for fusion materials at Oak Ridge National Laboratory (ORNL) uses an infrared plasma-arc lamp (PAL) to deliver incident heat fluxes as high as 27 MW/m2. The facility is being used to test irradiated plasma-facing component materials as part of the joint US-Japan PHENIX program. The irradiated samples are to be mounted on molybdenum sample holders attached to a water-cooled copper rod. Depending on the size and geometry of samples, several sample holders and copper rod configurations have been fabricated and tested. As a part of the effort to design sample holders compatible with the high heat flux (HHF) testing to be conducted at the IMTS facility, numerical simulations have been performed for two different water-cooled sample holder designs using the ANSYS FLUENT 14.0 commercial computational fluid dynamics (CFD) software package. The primary objective of this work is to evaluate the cooling capability of different sample holder designs, i.e. to estimate their maximum allowable incident heat flux values. 2D axisymmetric numerical simulations are performed using the realizable k-ε turbulence model and the RPI nucleate boiling model within ANSYS FLUENT 14.0. The results of the numerical model were compared against the experimental data for two sample holder designs tested in the IMTS facility. The model has been used to parametrically evaluate the effect of various operational parameters on the predicted temperature distributions. The results were used to identify the limiting parameter for safe operation of the two sample holders and the associated peak heat flux limits. The results of this investigation will help guide the development of new sample holder designs.

  5. Evaluation of Cooling Conditions for a High Heat Flux Testing Facility Based on Plasma-Arc Lamps

    DOE PAGES

    Charry, Carlos H.; Abdel-khalik, Said I.; Yoda, Minami; ...

    2015-07-31

    The new Irradiated Material Target Station (IMTS) facility for fusion materials at Oak Ridge National Laboratory (ORNL) uses an infrared plasma-arc lamp (PAL) to deliver incident heat fluxes as high as 27 MW/m2. The facility is being used to test irradiated plasma-facing component materials as part of the joint US-Japan PHENIX program. The irradiated samples are to be mounted on molybdenum sample holders attached to a water-cooled copper rod. Depending on the size and geometry of samples, several sample holders and copper rod configurations have been fabricated and tested. As a part of the effort to design sample holders compatiblemore » with the high heat flux (HHF) testing to be conducted at the IMTS facility, numerical simulations have been performed for two different water-cooled sample holder designs using the ANSYS FLUENT 14.0 commercial computational fluid dynamics (CFD) software package. The primary objective of this work is to evaluate the cooling capability of different sample holder designs, i.e. to estimate their maximum allowable incident heat flux values. 2D axisymmetric numerical simulations are performed using the realizable k-ε turbulence model and the RPI nucleate boiling model within ANSYS FLUENT 14.0. The results of the numerical model were compared against the experimental data for two sample holder designs tested in the IMTS facility. The model has been used to parametrically evaluate the effect of various operational parameters on the predicted temperature distributions. The results were used to identify the limiting parameter for safe operation of the two sample holders and the associated peak heat flux limits. The results of this investigation will help guide the development of new sample holder designs.« less

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

    SciTech Connect

    Morris, Robert N.; Baldwin, Charles A.; Demkowicz, Paul A.; Hunn, John D.; Reber, Edward L.

    2016-05-18

    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 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 failed 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. As a result, 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

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

    DOE PAGES

    Morris, Robert N.; Baldwin, Charles A.; Demkowicz, Paul A.; ...

    2016-05-18

    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 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium,more » 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 failed 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. As a result, 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

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

    PubMed

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

    2015-10-09

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

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

    PubMed Central

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

    2015-01-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-10-01

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

  11. Five years operating experience at the Fast Flux Test Facility

    SciTech Connect

    Baumhardt, R. J.; Bechtold, R. A.

    1987-04-01

    The Fast Flux Test Facility (FFTF) is a 400 Mw(t), loop-type, sodium-cooled, fast neutron reactor. It is operated by the Westinghouse Hanford Company for the United States Department of Energy at Richland, Washington. The FFTF is a multipurpose test reactor used to irradiate fuels and materials for programs such as Liquid Metal Reactor (LMR) research, fusion research, space power systems, isotope production and international research. FFTF is also used for testing concepts to be used in Advanced Reactors which will be designed to maximize passive safety features and not require complex shutdown systems to assure safe shutdown and heat removal. The FFTF also provides experience in the operation and maintenance of a reactor having prototypic components and systems typical of large LMR (LMFBR) power plants. The 5 year operational performance of the FFTF reactor is discussed in this report. 6 refs., 10 figs., 2 tabs.

  12. Reactor Simulator Testing Overview

    NASA Technical Reports Server (NTRS)

    Schoenfeld, Michael P.

    2013-01-01

    OBJECTIVE: Integrated testing of the TDU components TESTING SUMMARY: a) Verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. b) Thermal test heat regeneration design aspect of a cold trap purification filter. c) Pump performance test at pump voltages up to 150 V (targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V). TESTING HIGHLIGHTS: a) Gas and vacuum ground support test equipment performed effectively for NaK fill, loop pressurization, and NaK drain operations. b) Instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. c) 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. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  13. Tests of local transport theory and reduced wall impurity influx with highly radiative plasmas in the Tokamak Fusion Test Reactor

    NASA Astrophysics Data System (ADS)

    Hill, K. W.; Scott, S. D.; Bell, M.; Budny, R.; Bush, C. E.; Clark, R. E. H.; Denne-Hinnov, B.; Ernst, D. R.; Hammett, G. W.; Mikkelsen, D. R.; Mueller, D.; Ongena, J.; Park, H. K.; Ramsey, A. T.; Synakowski, E. J.; Taylor, G.; Zarnstorff, M. C.

    1999-03-01

    The electron temperature (Te) profile in neutral beam-heated supershot plasmas (Te0˜6-7 keV ion temperature Ti0˜15-20 keV, beam power Pb˜16 MW) was remarkably invariant when radiative losses were increased significantly through gas puffing of krypton and xenon in the Tokamak Fusion Test Reactor [McGuire et al., Phys. Plasmas 2, 2176 (1995)]. Trace impurity concentrations (nz/ne˜10-3) generated almost flat and centrally peaked radiation profiles, respectively, and increased the radiative losses to 45%-90% of the input power (from the normal ˜25%). Energy confinement was not degraded at radiated power fractions up to 80%. A 20%-30% increase in Ti, in spite of an increase in ion-electron power loss, implies a factor of ˜3 drop in the local ion thermal diffusivity. These experiments form the basis for a nearly ideal test of transport theory, since the change in the beam heating power profile is modest, while the distribution of power flow between (1) radiation and (2) conduction plus convection changes radically and is locally measurable. The decrease in Te was significantly less than predicted by two transport models and may provide important tests of more complete transport models. At input power levels of 30 MW, the increased radiation eliminated the catastrophic carbon influx (carbon "bloom") and performance (energy confinement and neutron production) was improved significantly relative to that of matched shots without impurity gas puffing.

  14. Recent view to the results of pulse tests in the IGR reactor with high burn-up fuel

    SciTech Connect

    Asmolov, V.; Yegorova, L.

    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.

  15. On the kinetics of the aluminum-water reaction during exposure in high-heat flux test loops: 1, A computer program for oxidation calculations

    SciTech Connect

    Pawel, R.E.

    1988-01-01

    The ''Griess Correlation,'' in which the thickness of the corrosion product on aluminum alloy surfaces is expressed as a function of time and temperature for high-flux-reactor conditions, was rewritten in the form of a simple, general rate equation. Based on this equation, a computer program that calculates oxide-layer thickness for any given time-temperature transient was written. 4 refs.

  16. ETR, TRA642. REACTOR FLOOR. EXPERIMENTAL GEAR AND TEST LOOP APPARATUS ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    ETR, TRA-642. REACTOR FLOOR. EXPERIMENTAL GEAR AND TEST LOOP APPARATUS SURROUND ETR. A WORKER STANDS OVER STORAGE POOL AND MANIPULATES CONTENTS WITH A LONG-HANDLED TOOL. INL NEGATIVE NO. 59-6015. Unknown Photographer, ca. 1959 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  17. Neutron flux profile monitor for use in a fission reactor

    DOEpatents

    Kopp, Manfred K.; Valentine, Kenneth H.

    1983-01-01

    A neutron flux monitor is provided which consists of a plurality of fission counters arranged as spaced-apart point detectors along a delay line. As a fission event occurs in any one of the counters, two delayed current pulses are generated at the output of the delay line. The time separation of the pulses identifies the counter in which the particular fission event occured. Neutron flux profiles of reactor cores can be more accurately measured as a result.

  18. Capabilities and Facilities Available at the Advanced Test Reactor to Support Development of the Next Generation Reactors

    SciTech Connect

    S. Blaine Grover; Raymond V. Furstenau

    2005-10-01

    The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. It is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The Irradiation Test Vehicle (ITV) installed in 1999 enhanced these capabilities by providing a built in experiment monitoring and control system for instrumented and/or temperature controlled experiments. This built in control system significantly reduces the cost for an actively monitored/temperature controlled experiments by providing the thermocouple connections, temperature control system, and temperature control gas supply and exhaust systems already in place at the irradiation position. Although the ITV in-core hardware was removed from the ATR during the last core replacement completed in early 2005, it (or a similar facility) could be re-installed for an irradiation program when the need arises. The proposed Gas Test Loop currently being designed for installation in the ATR will provide additional capability for testing of not only gas reactor materials and fuels but will also include enhanced fast flux rates for testing of materials and fuels for other next generation reactors including preliminary testing for fast reactor fuels and materials. This paper discusses the different irradiation capabilities available and the cost benefit issues related to each capability.

  19. High solids fermentation reactor

    DOEpatents

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

    1993-01-01

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

  20. High solids fermentation reactor

    DOEpatents

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

    1993-03-02

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

  1. Advanced Test Reactor - A National Scientific User Facility

    SciTech Connect

    Clifford J. Stanley

    2008-05-01

    The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected nuclear research reactor with a maximum operating power of 250 MWth. The unique serpentine configuration of the fuel elements creates five main reactor power lobes (regions) and nine flux traps. In addition to these nine flux traps there are 68 additional irradiation positions in the reactor core reflector tank. There are also 34 low-flux irradiation positions in the irradiation tanks outside the core reflector tank. The ATR is designed to provide a test environment for the evaluation of the effects of intense radiation (neutron and gamma). Due to the unique serpentine core design each of the five lobes can be operated at different powers and controlled independently. Options exist for the individual test trains and assemblies to be either cooled by the ATR coolant (i.e., exposed to ATR coolant flow rates, pressures, temperatures, and neutron flux) or to be installed in their own independent test loops where such parameters as temperature, pressure, flow rate, neutron flux, and energy can be controlled per experimenter specifications. The full-power maximum thermal neutron flux is ~1.0 x1015 n/cm2-sec with a maximum fast flux of ~5.0 x1014 n/cm2-sec. The Advanced Test Reactor, now a National Scientific User Facility, is a versatile tool in which a variety of nuclear reactor, nuclear physics, reactor fuel, and structural material irradiation experiments can be conducted. The cumulative effects of years of irradiation in a normal power reactor can be duplicated in a few weeks or months in the ATR due to its unique design, power density, and operating flexibility.

  2. High-Speed Tests of a Model Twin-Engine Low-Wing Transport Airplane

    NASA Technical Reports Server (NTRS)

    Becker, John V; LEONARD LLOYD H

    1942-01-01

    Report presents the results of force tests made of a 1/8-scale model of a twin-engine low-wing transport airplane in the NACA 8-foot high-speed tunnel to investigate compressibility and interference effects of speeds up to 450 miles per hour. In addition to tests of the standard arrangement of the model, tests were made with several modifications designed to reduce the drag and to increase the critical speed.

  3. A test fixture for measuring high-temperature hypersonic-engine seal performance

    NASA Technical Reports Server (NTRS)

    Steinetz, Bruce M.

    1990-01-01

    A test fixture for measuring the performance of several high temperature engine seal concepts was installed at the NASA Lewis Research Center. The test fixture was developed to evaluate seal concepts under development for advanced hypersonic engines such as those being considered for the National Aerospace Plane. The fixture can measure static seal leakage performance from room temperature up to 1500 F and air pressure differentials up to 100 psi. Performance of the seals can be measured while sealing against flat or engine simulated distorted walls, where distortions can be as large as 0.150 in. in only an 18 in. span. The fixture is designed to evaluate seals 3 feet long, a typical engine panel length. The seal channel can be configured to test square, circular, or rectangular seals that are nominally 0.5 in. high. The sensitivity of leakage performance to lateral or axial loading can also be measured using specially designed high temperature lateral and axial bellows preload systems. Leakage data for a candidate ceramic wafer engine seal is provided by way of example to demonstrate the test fixture's capabilities.

  4. Manufacturing and High Heat Flux Testing of Brazed Flat-Type W/CuCrZr Plasma Facing Components

    NASA Astrophysics Data System (ADS)

    Lian, Youyun; Liu, Xiang; Feng, Fan; Chen, Lei; Cheng, Zhengkui; Wang, Jin; Chen, Jiming

    2016-02-01

    Water-cooled flat-type W/CuCrZr plasma facing components with an interlayer of oxygen-free copper (OFC) have been developed by using vacuum brazing route. The OFC layer for the accommodation of thermal stresses was cast onto the surface of W at a temperature range of 1150 °C-1200 °C in a vacuum furnace. The W/OFC cast tiles were vacuum brazed to a CuCrZr heat sink at 940 °C using the silver-free filler material CuMnSiCr. The microstructure, bonding strength, and high heat flux properties of the brazed W/CuCrZr joint samples were investigated. The W/Cu joint exhibits an average tensile strength of 134 MPa, which is about the same strength as pure annealed copper. High heat flux tests were performed in the electron beam facility EMS-60. Experimental results indicated that the brazed W/CuCrZr mock-up experienced screening tests of up to 15 MW/m2 and cyclic tests of 9 MW/m2 for 1000 cycles without visible damage. supported by National Natural Science Foundation of China (No. 11205049) and the National Magnetic Confinement Fusion Science Program of China (No. 2011GB110004)

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

    SciTech Connect

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

    2014-01-01

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

  6. Application of High Speed Digital Image Correlation in Rocket Engine Hot Fire Testing

    NASA Technical Reports Server (NTRS)

    Gradl, Paul R.; Schmidt, Tim

    2016-01-01

    Hot fire testing of rocket engine components and rocket engine systems is a critical aspect of the development process to understand performance, reliability and system interactions. Ground testing provides the opportunity for highly instrumented development testing to validate analytical model predictions and determine necessary design changes and process improvements. To properly obtain discrete measurements for model validation, instrumentation must survive in the highly dynamic and extreme temperature application of hot fire testing. Digital Image Correlation has been investigated and being evaluated as a technique to augment traditional instrumentation during component and engine testing providing further data for additional performance improvements and cost savings. The feasibility of digital image correlation techniques were demonstrated in subscale and full scale hotfire testing. This incorporated a pair of high speed cameras to measure three-dimensional, real-time displacements and strains installed and operated under the extreme environments present on the test stand. The development process, setup and calibrations, data collection, hotfire test data collection and post-test analysis and results are presented in this paper.

  7. MAGNET ENGINEERING AND TEST RESULTS OF THE HIGH FIELD MAGNET R AND D PROGRAM AT BNL.

    SciTech Connect

    COZZOLINO,J.; ANERELLA,M.; ESCALLIER,J.; GANETIS,G.; GHOSH,A.; GUPTA,R.; HARRISON,M.; JAIN,A.; MARONE,A.; MURATORE,J.; PARKER,B.; SAMPSON,W.; SOIKA,R.; WANDERER,P.

    2002-08-04

    The Superconducting Magnet Division at Brookhaven National Laboratory (BNL) has been carrying out design, engineering, and technology development of high performance magnets for future accelerators. High Temperature Superconductors (HTS) play a major role in the BNL vision of a few high performance interaction region (IR) magnets that would be placed in a machine about ten years from now. This paper presents the engineering design of a ''react and wind'' Nb{sub 3}Sn magnet that will provide a 12 Tesla background field on HTS coils. In addition, the coil production tooling as well as the most recent 10-turn R&D coil test results will be discussed.

  8. ON-SITE ENGINEERING REPORT OF THE SLURRY-PHASE BIOLOGICAL REACTOR FOR PILOT-SCALE TESTING ON CONTAMINATED SOIL

    EPA Science Inventory

    The performance of pilot-scale bioslurry treatment on creosote-contaminated soil was evaluated. Five reactors containing 66 L of slurry (30% soil by weight), were operated in parallel. The soil was a sandy soil with minor gravel content. The pilot-scale phase utilized an inoculum...

  9. High-temperature test facility at the NASA Lewis engine components research laboratory

    NASA Technical Reports Server (NTRS)

    Colantonio, Renato O.

    1990-01-01

    The high temperature test facility (HTTF) at NASA-Lewis Engine Components Research Laboratory (ECRL) is presently used to evaluate the survivability of aerospace materials and the effectiveness of new sensing instrumentation in a realistic afterburner environment. The HTTF has also been used for advanced heat transfer studies on aerospace components. The research rig uses pressurized air which is heated with two combustors to simulate high temperature flow conditions for test specimens. Maximum airflow is 31 pps. The HTTF is pressure rated for up to 150 psig. Combustors are used to regulate test specimen temperatures up to 2500 F. Generic test sections are available to house test plates and advanced instrumentation. Customized test sections can be fabricated for programs requiring specialized features and functions. The high temperature test facility provides government and industry with a facility for testing aerospace components. Its operation and capabilities are described.

  10. Neutron flux reduction programs for reactor pressure vessel

    SciTech Connect

    Yoo, C.S.; Kim, B.C.

    2011-07-01

    The objective of this work is to implement various fast neutron flux reduction programs on the belt-line region of the reactor pressure vessel to reduce the increasing rate of reference temperature for pressurized thermal shock (RT PTS) for Korea Nuclear Unit 1. A pressurized thermal shock (PTS) event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in the belt-line region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Generally, the RT PTS value is continuously increasing according to the fast neutron irradiation during the reactor operation, and it can reach the screening criterion prior to the expiration of the operating license. To reduce the increasing rate of RT PTS, various neutron flux reduction programs can be implemented, which are focused on license renewal. In this paper, neutron flux reduction programs, such as low leakage loading pattern strategy, loading of neutron absorber rods, and dummy fuel assembly loading are considered for Korea Nuclear Unit 1, of which the RT PTS value of the leading material (circumferential weld) is going to reach the screening criterion in the near future. To evaluate the effects of the neutron flux reduction programs, plant and cycle specific forward neutron transport calculations for the various neutron flux reduction programs were carried out. For the analysis, all transport calculations were carried out by using the DORT 3.1 discrete ordinate code and BUGLE-96 cross-section library. (authors)

  11. Transpiring wall supercritical water oxidation test reactor design report

    SciTech Connect

    Haroldsen, B.L.; Ariizumi, D.Y.; Mills, B.E.; Brown, B.G.; Rousar, D.C.

    1996-02-01

    Sandia National Laboratories is working with GenCorp, Aerojet and Foster Wheeler Development Corporation to develop a transpiring wall supercritical water oxidation reactor. The transpiring wall reactor promises to mitigate problems of salt deposition and corrosion by forming a protective boundary layer of pure supercritical water. A laboratory scale test reactor has been assembled to demonstrate the concept. A 1/4 scale transpiring wall reactor was designed and fabricated by Aerojet using their platelet technology. Sandia`s Engineering Evaluation Reactor serves as a test bed to supply, pressurize and heat the waste; collect, measure and analyze the effluent; and control operation of the system. This report describes the design, test capabilities, and operation of this versatile and unique test system with the transpiring wall reactor.

  12. Reactor Antineutrino Flux and Spectrum Shape from Daya Bay

    NASA Astrophysics Data System (ADS)

    Napolitano, Jim; Daya Bay Collaboration

    2017-01-01

    The Daya Bay Reactor Neutrino Experiment has collected very large samples of νe p ->e+ n events, where the νe are from the cores of six power plant reactors that undergo regular refueling. With 621 days of data, more than 1.2 million events of this type were detected. The collaboration has analyzed these data in terms of the absolute flux (addressing the ``Reactor Neutrino Anomaly''), the spectrum shape (including the excess in the region of 5 MeV prompt energy), and other effects. This talk will summarize the results from our most recent analyses, and discuss new initiatives aimed at continuing to understand the fine detail of the reactor νe spectrum.

  13. Jet engine testing apparatus

    SciTech Connect

    Zweifel, T.L.

    1987-03-24

    An apparatus is described for testing jet engines mounted on an aircraft, the jet engines of the type having a high speed rotor and a low speed rotor, comprising: representative signal means for providing first representative signals representative of rotation rates of the low speed rotor in the jet engines and second representative signals representative of rotation rates of the high speed rotor in the jet engines; equivalent signal means coupled to receive the second representative signals for deriving equivalent signals representative of low speed rotor rotation rates of normally operating jet engines having high speed rotor rotation rates represented by the second representative signals; first difference signal means coupled to receive the first representative signals and the equivalent signals for providing first difference signals representative of differences between the first representative signals and the equivalent signals; means for providing threshold signals; first detector means coupled to the threshold signal means and the first difference signal means for comparing the threshold signals and the first difference signals to provide first detected signals representative of values of the first difference signals relative to the threshold signals; and engine failure indicator means coupled to receive the detected signals for determination of engine failures.

  14. In-reactor testing of the closed cycle gas core reactor: The Nuclear Light Bulb concept

    NASA Astrophysics Data System (ADS)

    Gauntt, R. O.; Slutz, S. A.; Harms, G. A.; Latham, T. S.; Roman, W. C.; Rodgers, R. J.

    1992-10-01

    The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (greater than 1800 s) and thrust (greater than 445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (greater than 4000 K). Analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept are described. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRE. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRE for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept.

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

    SciTech Connect

    Laurie, M.; Fourrez, S.; Fuetterer, M. A.; Lapetite, J. M.

    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)

  16. Space Electronic Test Engineering

    NASA Technical Reports Server (NTRS)

    Chambers, Rodney D.

    2004-01-01

    The Space Power and Propulsion Test Engineering Branch at NASA Glenn Research center has the important duty of controlling electronic test engineering services. These services include test planning and early assessment of Space projects, management and/or technical support required to safely and effectively prepare the article and facility for testing, operation of test facilities, and validation/delivery of data to customer. The Space Electronic Test Engineering Branch is assigned electronic test engineering responsibility for the GRC Space Simulation, Microgravity, Cryogenic, and Combustion Test Facilities. While working with the Space Power and Propulsion Test Engineering Branch I am working on several different assignments. My primary assignment deals with an electrical hardware unit known as Sunny Boy. Sunny Boy is a DC load Bank that is designed for solar arrays in which it is used to convert DC power form the solar arrays into AC power at 60 hertz to pump back into the electricity grid. However, there are some researchers who decided that they would like to use the Sunny Boy unit in a space simulation as a DC load bank for a space shuttle or even the International Space Station hardware. In order to do so I must create a communication link between a computer and the Sunny Boy unit so that I can preset a few of the limits (such power, set & constant voltage levels) that Sunny Boy will need to operate using the applied DC load. Apart from this assignment I am also working on a hi-tech circuit that I need to have built at a researcher s request. This is a high voltage analog to digital circuit that will be used to record data from space ion propulsion rocket booster tests. The problem that makes building this circuit so difficult is that it contains high voltage we must find a way to lower the voltage signal before the data is transferred into the computer to be read. The solution to this problem was to transport the signal using infrared light which will lower

  17. Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect

    E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

    2003-10-28

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

  18. Laser High-Cycle Thermal Fatigue of Pulse Detonation Engine Combustor Materials Tested

    NASA Technical Reports Server (NTRS)

    Zhu, Dong-Ming; Fox, Dennis S.; Miller, Robert A.

    2001-01-01

    Pulse detonation engines (PDE's) have received increasing attention for future aerospace propulsion applications. Because the PDE is designed for a high-frequency, intermittent detonation combustion process, extremely high gas temperatures and pressures can be realized under the nearly constant-volume combustion environment. The PDE's can potentially achieve higher thermodynamic cycle efficiency and thrust density in comparison to traditional constant-pressure combustion gas turbine engines (ref. 1). However, the development of these engines requires robust design of the engine components that must endure harsh detonation environments. In particular, the detonation combustor chamber, which is designed to sustain and confine the detonation combustion process, will experience high pressure and temperature pulses with very short durations (refs. 2 and 3). Therefore, it is of great importance to evaluate PDE combustor materials and components under simulated engine temperatures and stress conditions in the laboratory. In this study, a high-cycle thermal fatigue test rig was established at the NASA Glenn Research Center using a 1.5-kW CO2 laser. The high-power laser, operating in the pulsed mode, can be controlled at various pulse energy levels and waveform distributions. The enhanced laser pulses can be used to mimic the time-dependent temperature and pressure waves encountered in a pulsed detonation engine. Under the enhanced laser pulse condition, a maximum 7.5-kW peak power with a duration of approximately 0.1 to 0.2 msec (a spike) can be achieved, followed by a plateau region that has about one-fifth of the maximum power level with several milliseconds duration. The laser thermal fatigue rig has also been developed to adopt flat and rotating tubular specimen configurations for the simulated engine tests. More sophisticated laser optic systems can be used to simulate the spatial distributions of the temperature and shock waves in the engine. Pulse laser high

  19. Conceptual Design of Vacuum Chamber for testing of high heat flux components using electron beam as a source

    NASA Astrophysics Data System (ADS)

    Khan, M. S.; Swamy, Rajamannar; Khirwadkar, S. S.; Divertors Division, Prototype

    2012-11-01

    A conceptual design of vacuum chamber is proposed to study the thermal response of high heat flux components under energy depositions of the magnitude and durations expected in plasma fusion devices. It is equipped with high power electron beam with maximum beam power of 200 KW mounted in a stationary horizontal position from back side of the chamber. The electron beam is used as a heat source to evaluate the heat removal capacity, material performance under thermal loads & stresses, thermal fatigue etc on actively cooled mock - ups which are mounted on a flange system which is the front side door of the chamber. The tests mock - ups are connected to a high pressure high temperature water circulation system (HPHT-WCS) operated over a wide range of conditions. The vacuum chamber consists of different ports at different angles to view the mock -up surface available for mock -up diagnostics. The vacuum chamber is pumped with different pumps mounted on side ports of the chamber. The chamber is shielded from X - rays which are generated inside the chamber when high-energy electrons are incident on the mock-up. The design includes development of a conceptual design with theoretical calculations and CAD modelling of the system using CATIA V5. These CAD models give an outline on the complete geometry of HHF test chamber, fabrication challenges and safety issues. FEA analysis of the system has been performed to check the structural integrity when the system is subjected to structural & thermal loads.

  20. Liquid Rocket Engine Testing

    DTIC Science & Technology

    2016-10-21

    Briefing Charts 3. DATES COVERED (From - To) 17 October 2016 – 26 October 2016 4. TITLE AND SUBTITLE Liquid Rocket Engine Testing 5a. CONTRACT NUMBER...298 (Rev. 8-98) Prescribed by ANSI Std. 239.18 Liquid Rocket Engine Testing SFTE Symposium 21 October 2016 Jake Robertson, Capt USAF AFRL...Distribution Unlimited. PA Clearance 16493 Liquid Rocket Engine TestingEngines and their components are extensively static- tested in development • This

  1. Liquid rocket engine test facility engineering challenges

    NASA Astrophysics Data System (ADS)

    Ellerbrock, Hartwig; Ziegenhagen, Stefan

    2006-12-01

    Liquid rocket engines for launch vehicles and space crafts as well as their subsystems need to be verified and qualified during hot-runs. A high test cadence combined with a flexible test team helps to reduce the cost for test verification during development/qualification as well as during acceptance testing for production. Test facility intelligence allows to test subsystems in the same manner as during complete engine system tests and will therefore reduce development time and cost. This paper gives an overview of the maturing of test engineering know how for rocket engine test stands as well as high altitude test stands for small propulsion thrusters at EADS-ST in Ottobrunn and Lampoldshausen and is split into two parts: Part 1 gives a historical overview of the EADS-ST test stands at Ottobrunn and Lampoldshausen since the beginning of Rocket propulsion activities in the 1960s. Part 2 gives an overview of the actual test capabilities and the test engineering know-how for test stand construction/adaptation and their use during running programs. Examples of actual realised facility concepts are given to demonstrate cost saving potential for test programs in both cases for development/qualification issues as well as for production purposes.

  2. Advanced Test Reactor National Scientific User Facility Progress

    SciTech Connect

    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

  3. Combined analysis of neutron and photon flux measurements for the Jules Horowitz reactor core mapping

    SciTech Connect

    Fourmentel, D.; Villard, J. F.; Lyoussi, A.; Reynard-Carette, C.; Bignan, G.; Chauvin, J. P.; Gonnier, C.; Guimbal, P.; Malo, J. Y.; Carette, M.; Janulyte, A.; Merroun, O.; Brun, J.; Zerega, Y.; Andre, J.

    2011-07-01

    We study the combined analysis of nuclear measurements to improve the knowledge of the irradiation conditions in the experimental locations of the future Jules Horowitz Reactor (JHR). The goal of the present work is to measure more accurately neutron flux, photon flux and nuclear heating in the reactor. In a Material Testing Reactor (MTR), nuclear heating is a crucial parameter to design the experimental devices to be irradiated in harsh nuclear conditions. This parameter drives the temperature of the devices and of the samples. The numerical codes can predict this parameter but in-situ measurements are necessary to reach the expected accuracy. For this reason, one objective of the IN-CORE program [1] is to study the combined measurements of neutron and photon flux and their cross advanced interpretation. It should be reminded that both neutron and photon sensors are not totally selective as their signals are due to neutron and photon interactions. We intend to measure the neutron flux by three different kinds of sensors (Uranium Fission chamber, Plutonium Fission chamber and Self Powered Neutron Detector), the photon flux by two different sensors (Ionization chamber and Self Powered Gamma Detector) and the nuclear heating by two different ones (Differential calorimeter and Gamma Thermometer). For the same parameter, we expect that the use of different kinds of sensors will allow a better estimation of the aimed parameter by mixing different spectrum responses and different neutron and gamma contributions. An experimental test called CARMEN-1 is scheduled in OSIRIS reactor (CEA Saclay - France) at the end of 2011, with the goal to map irradiation locations in the reactor reflector to get a first validation of the analysis model. This article focuses on the sensor selection for CARMEN-1 experiment and to the way to link neutron and photon flux measurements in view to reduce their uncertainties but also to better assess the neutron and photon contributions to nuclear

  4. Nuclear Island Engineering MHTGR [Modular High-Temperature Gas-cooled Reactor] preliminary and final designs. Technical progress report, December 12, 1988--September 30, 1989

    SciTech Connect

    1989-12-01

    This report summarizes the Department of Energy (DOE)-funded work performed by General Atomics (GA) under the Nuclear Island Engineering (NIE)-Modular High-Temperature Gas-cooled Reactor (MHTGR) Preliminary and Final Designs Contract DE-AC03-89SF17885 for the period December 12, 1988 through September 30, 1989. This reporting period is the first (partial) fiscal year of the 5-year contract performance period. The objective of DOE`s MHTGR program is to advance the design from the conceptual design phase into preliminary design and then on to final design in support of the Nuclear Regulatory Commission`s (NRC`s) design review and approval of the MHTGR Design Team, is focused on the Nuclear Island portion of the technology and design, primarily in the areas of the reactor and internals, fuel characteristics and fuel fabrication, helium services systems, reactor protection, shutdown cooling, circulator design, and refueling system. Maintenance and implementation of the functional methodology, plant-level analysis, support for probabilistic risk assessment, quality assurance, operations, and reliability/availability assessments are included in GA`s scope of work.

  5. High Performance Arcjet Engines

    NASA Technical Reports Server (NTRS)

    Kennel, Elliot B.; Ivanov, Alexey Nikolayevich; Nikolayev, Yuri Vyacheslavovich

    1994-01-01

    This effort sought to exploit advanced single crystal tungsten-tantalum alloy material for fabrication of a high strength, high temperature arcjet anode. The use of this material is expected to result in improved strength, temperature resistance, and lifetime compared to state of the art polycrystalline alloys. In addition, the use of high electrical and thermal conductivity carbon-carbon composites was considered, and is believed to be a feasible approach. Highly conductive carbon-carbon composite anode capability represents enabling technology for rotating-arc designs derived from the Russian Scientific Research Institute of Thermal Processes (NIITP) because of high heat fluxes at the anode surface. However, for US designs the anode heat flux is much smaller, and thus the benefits are not as great as in the case of NIITP-derived designs. Still, it does appear that the tensile properties of carbon-carbon can be even better than those of single crystal tungsten alloys, especially when nearly-single-crystal fibers such as vapor grown carbon fiber (VGCF) are used. Composites fabricated from such materials must be coated with a refractory carbide coating in order to ensure compatibility with high temperature hydrogen. Fabrication of tungsten alloy single crystals in the sizes required for fabrication of an arcjet anode has been shown to be feasible. Test data indicate that the material can be expected to be at least the equal of W-Re-HfC polycrystalline alloy in terms of its tensile properties, and possibly superior. We are also informed by our colleagues at Scientific Production Association Luch (NP0 Luch) that it is possible to use Russian technology to fabricate polycrystalline W-Re-HfC or other high strength alloys if desired. This is important because existing engines must rely on previously accumulated stocks of these materials, and a fabrication capability for future requirements is not assured.

  6. Space reactor fuel element testing in upgraded TREAT

    SciTech Connect

    Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

    1993-01-14

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.

  7. Space reactor fuel element testing in upgraded TREAT

    SciTech Connect

    Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

    1993-05-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.

  8. Performance and Environmental Test Results of the High Voltage Hall Accelerator Engineering Development Unit

    NASA Technical Reports Server (NTRS)

    Kamhawi, Hani; Haag, Thomas; Huang, Wensheng; Shastry, Rohit; Pinero, Luis; Peterson, Todd; Mathers, Alex

    2012-01-01

    NASA Science Mission Directorate's In-Space Propulsion Technology Program is sponsoring the development of a 3.5 kW-class engineering development unit Hall thruster for implementation in NASA science and exploration missions. NASA Glenn and Aerojet are developing a high fidelity high voltage Hall accelerator that can achieve specific impulse magnitudes greater than 2,700 seconds and xenon throughput capability in excess of 300 kilograms. Performance, plume mappings, thermal characterization, and vibration tests of the high voltage Hall accelerator engineering development unit have been performed. Performance test results indicated that at 3.9 kW the thruster achieved a total thrust efficiency and specific impulse of 58%, and 2,700 sec, respectively. Thermal characterization tests indicated that the thruster component temperatures were within the prescribed material maximum operating temperature limits during full power thruster operation. Finally, thruster vibration tests indicated that the thruster survived the 3-axes qualification full-level random vibration test series. Pre and post-vibration test performance mappings indicated almost identical thruster performance. Finally, an update on the development progress of a power processing unit and a xenon feed system is provided.

  9. Introduction to Chemical Engineering Reactor Analysis: A Web-Based Reactor Design Game

    ERIC Educational Resources Information Center

    Orbey, Nese; Clay, Molly; Russell, T.W. Fraser

    2014-01-01

    An approach to explain chemical engineering through a Web-based interactive game design was developed and used with college freshman and junior/senior high school students. The goal of this approach was to demonstrate how to model a lab-scale experiment, and use the results to design and operate a chemical reactor. The game incorporates both…

  10. Energy Efficient Engine: High-pressure compressor test hardware detailed design report

    NASA Technical Reports Server (NTRS)

    Howe, David C.; Marchant, R. D.

    1988-01-01

    The objective of the NASA Energy Efficient Engine program is to identify and verify the technology required to achieve significant reductions in fuel consumption and operating cost for future commercial gas turbine engines. The design and analysis is documented of the high pressure compressor which was tested as part of the Pratt and Whitney effort under the Energy Efficient Engine program. This compressor was designed to produce a 14:1 pressure ratio in ten stages with an adiabatic efficiency of 88.2 percent in the flight propulsion system. The corresponding expected efficiency for the compressor component test rig is 86.5 percent. Other performance goals are a surge margin of 20 percent, a corrected flow rate of 35.2 kg/sec (77.5 lb/sec), and a life of 20,000 missions and 30,000 hours. Low loss, highly loaded airfoils are used to increase efficiency while reducing the parts count. Active clearance control and case trenches in abradable strips over the blade tips are included in the compressor component design to further increase the efficiency potential. The test rig incorporates variable geometry stator vanes in all stages to permit maximum flexibility in developing stage-to-stage matching. This provision precluded active clearance control on the rear case of the test rig. Both the component and rig designs meet or exceed design requirements with the exception of life goals, which will be achievable with planned advances in materials technology.

  11. Fast Flux Test Facility final safety analysis report. Amendment 72

    SciTech Connect

    Gantt, D. A.

    1992-08-01

    This document provides the Final Safety Analysis Report (FSAR) Amendment 72 for incorporation into the Fast Flux Test Facility (FFTF) FSAR set. This amendment change incorporates Engineering Change Notices issued subsequent to Amendment 71 and approved for incorporation before June 24, 1992. These include changes in: Chapter 2, Site Characteristics; Chapter 3, Design Criteria Structures, Equipment, and Systems; Chapter 5B, Reactor Coolant System; Chapter 7, Instrumentation and Control Systems; Chapter 8, Electrical Systems - The description of the Class 1E, 125 Vdc systems is updated for the higher capacity of the newly installed, replacement batteries; Chapter 9, Auxiliary Systems - The description of the inert cell NASA systems is corrected to list the correct number of spare sample points; Chapter 11, Reactor Refueling System; Chapter 12, Radiation Protection and Waste Management; Chapter 13, Conduct of Operations; Chapter 16, Quality Assurance; Chapter 17, Technical Specifications; Chapter 19, FFTF Fire Specifications for Fire Detection, Alarm, and Protection Systems; Chapter 20, FFTF Criticality Specifications; and Appendix B, Primary Piping Integrity Evaluation.

  12. Flight Test Engineering

    NASA Technical Reports Server (NTRS)

    Pavlock, Kate Maureen

    2013-01-01

    Although the scope of flight test engineering efforts may vary among organizations, all point to a common theme: flight test engineering is an interdisciplinary effort to test an asset in its operational flight environment. Upfront planning where design, implementation, and test efforts are clearly aligned with the flight test objective are keys to success. This chapter provides a top level perspective of flight test engineering for the non-expert. Additional research and reading on the topic is encouraged to develop a deeper understanding of specific considerations involved in each phase of flight test engineering.

  13. Reactor Simulator Integration and Testing

    NASA Technical Reports Server (NTRS)

    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.

  14. Safety philosophy of gas turbine high temperature reactor (GTHTR300)

    SciTech Connect

    Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa

    2002-07-01

    Japan Atomic Energy Research Institute (JAERI) has undertaken the study of an original design concept of gas turbine high temperature reactor, the GTHTR300. The general concept of this study is development of a greatly simplified design that leads to substantially reduced technical and cost requirements. Newly proposed design features enable the GTHTR300 to be an efficient and economically competitive reactor in 2010's. Also, the GTHTR300 fully takes advantage of its inherent safety characteristics. The safety philosophy of the GTHTR300 is developed based on the HTTR (High Temperature Engineering Test Reactor) of JAERI which is the first HTGR in Japan. Major features of the newly proposed safety philosophy for the GTHTR300 are described in this article. (authors)

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

    NASA Astrophysics Data System (ADS)

    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

  16. Calorimeter measures high nuclear heating rates and their gradients across a reactor test hole

    NASA Technical Reports Server (NTRS)

    Burwell, D.; Coombe, J. R.; Mc Bride, J.

    1970-01-01

    Pedestal-type calorimeter measures gamma-ray heating rates from 0.5 to 7.0 watts per gram of aluminum. Nuclear heating rate is a function of cylinder temperature change, measured by four chromel-alumel thermocouples attached to the calorimeter, and known thermoconductivity of the tested material.

  17. A Transient Numerical Simulation of Perched Ground-Water Flow at the Test Reactor Area, Idaho National Engineering and Environmental Laboratory, Idaho, 1952-94

    SciTech Connect

    B. R. Orr

    1999-11-01

    Studies of flow through the unsaturated zone and perched ground-water zones above the Snake River Plain aquifer are part of the overall assessment of ground-water flow and determination of the fate and transport of contaminants in the subsurface at the Idaho National Engineering and Environmental Laboratory (INEEL). These studies include definition of the hydrologic controls on the formation of perched ground-water zones and description of the transport and fate of wastewater constituents as they moved through the unsaturated zone. The definition of hydrologic controls requires stratigraphic correlation of basalt flows and sedimentary interbeds within the saturated zone, analysis of hydraulic properties of unsaturated-zone rocks, numerical modeling of the formation of perched ground-water zones, and batch and column experiments to determine rock-water geochemical processes. This report describes the development of a transient numerical simulation that was used to evaluate a conceptual model of flow through perched ground-water zones beneath wastewater infiltration ponds at the Test Reactor Area (TRA).

  18. Photographic documentation of the High Power Engine Propulsion HiPEP after a duration test. Also ph

    NASA Technical Reports Server (NTRS)

    2005-01-01

    Photographic documentation of the High Power Engine Propulsion HiPEP after a duration test. Also photographed are the instrumentation and installation articles to reveal post test conditions such as corrosion and pitting.

  19. Performance of AGR-1 High-Temperature Reactor Fuel During Post-Irradiation Heating Tests

    SciTech Connect

    Morris, Robert Noel; Baldwin, Charles A; Hunn, John D; Demkowicz, Paul; Reber, Edward

    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.

  20. Startup of the FFTF sodium cooled reactor. [Acceptance Test Program

    SciTech Connect

    Redekopp, R.D.; Umek, A.M.

    1981-03-01

    The Fast Flux Test Facility (FFTF), located on the Department of Energy (DOE) Hanford Reservation near Richland, Washington, is a 3 Loop 400 MW(t) sodium cooled fast reactor with a primary mission to test fuels and materials for development of the Liquid Metal Fast Breeder Reactor (LMFBR). Bringing FFTF to a condition to accomplish this mission is the goal of the Acceptance Test Program (ATP). This program was the mechanism for achieving startup of the FFTF. Highlights of the ATP involving the system inerting, liquid metal and inerted cell testing and initial ascent to full power are discussed.

  1. Energy efficient engine high-pressure turbine component rig performance test report

    NASA Technical Reports Server (NTRS)

    Leach, K. P.

    1983-01-01

    A rig test of the cooled high-pressure turbine component for the Energy Efficient Engine was successfully completed. The principal objective of this test was to substantiate the turbine design point performance as well as determine off-design performance with the interaction of the secondary flow system. The measured efficiency of the cooled turbine component was 88.5 percent, which surpassed the rig design goal of 86.5 percent. The secondary flow system in the turbine performed according to the design intent. Characterization studies showed that secondary flow system performance is insensitive to flow and pressure variations. Overall, this test has demonstrated that a highly-loaded, transonic, single-stage turbine can achieve a high level of operating efficiency.

  2. Design and Laboratory Evaluation of Future Elongation and Diameter Measurements at the Advanced Test Reactor

    SciTech Connect

    K. L. Davis; D. L. Knudson; J. L. Rempe; J. C. Crepeau; S. Solstad

    2015-07-01

    New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can undergo significant dimensional and physical changes during high temperature irradiations. In order to accurately predict these changes, real-time data must be obtained under prototypic irradiation conditions for model development and validation. To provide such data, researchers at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) are developing several instrumented test rigs to obtain data real-time from specimens irradiated in well-controlled pressurized water reactor (PWR) coolant conditions in the Advanced Test Reactor (ATR). This paper reports the status of INL efforts to develop and evaluate prototype test rigs that rely on Linear Variable Differential Transformers (LVDTs) in laboratory settings. Although similar LVDT-based test rigs have been deployed in lower flux Materials Testing Reactors (MTRs), this effort is unique because it relies on robust LVDTs that can withstand higher temperatures and higher fluxes than often found in other MTR irradiations. Specifically, the test rigs are designed for detecting changes in length and diameter of specimens irradiated in ATR PWR loops. Once implemented, these test rigs will provide ATR users with unique capabilities that are sorely needed to obtain measurements such as elongation caused by thermal expansion and/or creep loading and diameter changes associated with fuel and cladding swelling, pellet-clad interaction, and crud buildup.

  3. High-power baseline and motoring test results for the GPU-3 Stirling engine

    NASA Technical Reports Server (NTRS)

    Thieme, L. G.

    1981-01-01

    Test results are given for the full power range of the engine with both helium and hydrogen working fluids. Comparisons are made to previous testing using an alternator and resistance load bank to absorb the engine output. Indicated power results are presented as determined by several methods. Motoring tests were run to aid in determining engine mechanical losses. Comparisons are made between the results of motoring and energy-balance methods for finding mechanical losses.

  4. REACTOR FUEL ELEMENTS TESTING CONTAINER

    DOEpatents

    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)

  5. Development of a reactor engineering workstation at Seabrook station

    SciTech Connect

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

    1992-01-01

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

  6. Prototype thin-film thermocouple/heat-flux sensor for a ceramic-insulated diesel engine

    NASA Technical Reports Server (NTRS)

    Kim, Walter S.; Barrows, Richard F.

    1988-01-01

    A platinum versus platinum-13 percent rhodium thin-film thermocouple/heat-flux sensor was devised and tested in the harsh, high-temperature environment of a ceramic-insulated, low-heat-rejection diesel engine. The sensor probe assembly was developed to provide experimental validation of heat transfer and thermal analysis methodologies applicable to the insulated diesel engine concept. The thin-film thermocouple configuration was chosen to approximate an uninterrupted chamber surface and provide a 1-D heat-flux path through the probe body. The engine test was conducted by Purdue University for Integral Technologies, Inc., under a DOE-funded contract managed by NASA Lewis Research Center. The thin-film sensor performed reliably during 6 to 10 hr of repeated engine runs at indicated mean surface temperatures up to 950 K. However, the sensor suffered partial loss of adhesion in the thin-film thermocouple junction area following maximum cyclic temperature excursions to greater than 1150 K.

  7. Engine Test and Measurements

    NASA Technical Reports Server (NTRS)

    Wey, Chown Chou

    1999-01-01

    Although the importance of aerosols and their precursors are now well recognized, the characterization of current subsonic engines for these emissions is far from complete. Furthermore, since the relationship of engine operating parameters to aerosol emissions is not known, extrapolation to untested and unbuilt engines necessarily remains highly uncertain. 1997 NASA LaRC engine test, as well as the parallel 1997 NASA LaRC flight measurement, attempts to address both issues by expanding measurements of aerosols and aerosol precursors with fuels containing different levels of fuel sulfur content. The specific objective of the 1997 engine test is to obtain a database of sulfur oxides emissions as well as the non-volatile particulate emission properties as a function of fuel sulfur and engine operating conditions. Four diagnostic systems, extractive and non-intrusive (optical), will be assembled for the gaseous and particulate emissions characterization measurements study. NASA is responsible for the extractive gaseous emissions measurement system which contains an array of analyzers dedicated to examining the concentrations of specific gases (NO, NO(x), CO, CO2, O2, THC, SO2) and the smoke number. University of Missouri-Rolla uses the Mobile Aerosol Sampling System to measure aerosol/particulate total concentration, size distribution, volatility and hydration property. Air Force Research Laboratory uses the Chemical Ionization Mass Spectrometer to measure SO2, SO3/H2SO4, and HN03 Aerodyne Research, Inc. uses Infrared Tunable Diode Laser system to measure SO2, SO3, NO, H2O, and CO2.

  8. Solar-thermal engine testing

    NASA Astrophysics Data System (ADS)

    Tucker, Stephen; Salvail, Pat

    2002-01-01

    A solar-thermal engine serves as a high-temperature solar-radiation absorber, heat exchanger, and rocket nozzle, collecting concentrated solar radiation into an absorber cavity and transferring this energy to a propellant as heat. Propellant gas can be heated to temperatures approaching 4,500 °F and expanded in a rocket nozzle, creating low thrust with a high specific impulse (Isp). The Shooting Star Experiment (SSE) solar-thermal engine is made of 100 percent chemically vapor deposited (CVD) rhenium. The engine ``module'' consists of an engine assembly, propellant feedline, engine support structure, thermal insulation, and instrumentation. Engine thermal performance tests consist of a series of high-temperature thermal cycles intended to characterize the propulsive performance of the engines and the thermal effectiveness of the engine support structure and insulation system. A silicone-carbide electrical resistance heater, placed inside the inner shell, substitutes for solar radiation and heats the engine. Although the preferred propellant is hydrogen, the propellant used in these tests is gaseous nitrogen. Because rhenium oxidizes at elevated temperatures, the tests are performed in a vacuum chamber. Test data will include transient and steady state temperatures on selected engine surfaces, propellant pressures and flow rates, and engine thrust levels. The engine propellant-feed system is designed to supply GN2 to the engine at a constant inlet pressure of 60 psia, producing a near-constant thrust of 1.0 lb. Gaseous hydrogen will be used in subsequent tests. The propellant flow rate decreases with increasing propellant temperature, while maintaining constant thrust, increasing engine Isp. In conjunction with analytical models of the heat exchanger, the temperature data will provide insight into the effectiveness of the insulation system, the structural support system, and the overall engine performance. These tests also provide experience on operational aspects

  9. Beryllium Use in the Advanced Test Reactor

    SciTech Connect

    Glen R. Longhurst

    2007-12-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) began operation in 1967. It makes use of a unique serpentine fuel core design and a beryllium reflector. Reactor control is achieved with rotating beryllium cylinders to which have been fastened plates of hafnium. Over time, the beryllium develops rather high helium content because of nuclear transmutations and begins to swell. The beryllium must be replaced at nominally 10-year intervals. Determination of when the replacement is made is by visual observation using a periscope to examine the beryllium surface for cracking and swelling. Disposition of the irradiated beryllium was once accomplished in the INL’s Radioactive Waste Management Complex, but that is no longer possible. Among contributing reasons are high levels of specific radioactive contaminants including transuranics. The INL is presently considering disposition pathways for this irradiated beryllium, but presently is storing it in the canal adjacent to the reactor. Numerous issues are associated with this situation including (1) Is there a need for ultra-low uranium material? (2) Is there a need to recover tritium from irradiated beryllium either because this is a strategic material resource or in preparation for disposal? (3) Is there a need to remove activation and fission products from irradiated beryllium? (4) Will there be enough material available to meet requirements for research reactors (fission and fusion)? In this paper will be discussed the present status of considerations on these issues.

  10. An experimental study of external reactor vessel cooling strategy on the critical heat flux using the graphene oxide nano-fluid

    SciTech Connect

    Park, S. D.; Lee, S. W.; Kang, S.; Kim, S. M.; Seo, H.; Bang, I. C.

    2012-07-01

    External reactor vessel cooling (ERVC) for in-vessel retention (IVR) of corium as a key severe accident management strategy can be achieved by flooding the reactor cavity during a severe accident. In this accident mitigation strategy, the decay heat removal capability depends on whether the imposed heat flux exceeds critical heat flux (CHF). To provide sufficient cooling for high-power reactors such as APR1400, there have been some R and D efforts to use the reactor vessel with micro-porous coating and nano-fluids boiling-induced coating. The dispersion stability of graphene-oxide nano-fluid in the chemical conditions of flooding water that includes boric acid, lithium hydroxide (LiOH) and tri-sodium phosphate (TSP) was checked in terms of surface charge or zeta potential before the CHF experiments. Results showed that graphene-oxide nano-fluids were very stable under ERVC environment. The critical heat flux (CHF) on the reactor vessel external wall was measured using the small scale two-dimensional slide test section. The radius of the curvature is 0.1 m. The dimension of each part in the facility simulated the APR-1400. The heater was designed to produce the different heat flux. The magnitude of heat flux follows the one of the APR-1400 when the severe accident occurred. All tests were conducted under inlet subcooling 10 K. Graphene-oxide nano-fluids (concentration: 10 -4 V%) enhanced CHF limits up to about 20% at mass flux 50 kg/m{sup 2}s and 100 kg/m{sup 2}s in comparison with the results of the distilled water at same test condition. (authors)

  11. Recovery of Information from the Fast Flux Test Facility for the Advanced Fuel Cycle Initiative

    SciTech Connect

    Nielsen, Deborah L.; Makenas, Bruce J.; Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.

    2009-09-30

    The Fast Flux Test Facility is the most recent Liquid Metal Reactor to operate in the United States. Information from the design, construction, and operation of this reactor was at risk as the facilities associated with the reactor are being shut down. The Advanced Fuel Cycle Initiative is a program managed by the Office of Nuclear Energy of the U.S. Department of Energy with a mission to develop new fuel cycle technologies to support both current and advanced reactors. Securing and preserving the knowledge gained from operation and testing in the Fast Flux Test Facility is an important part of the Knowledge Preservation activity in this program.

  12. Development of advanced high-temperature heat flux sensors

    NASA Technical Reports Server (NTRS)

    Atkinson, W. H.; Strange, R. R.

    1982-01-01

    Various configurations of high temperature, heat flux sensors were studied to determine their suitability for use in experimental combustor liners of advanced aircraft gas turbine engines. It was determined that embedded thermocouple sensors, laminated sensors, and Gardon gauge sensors, were the most viable candidates. Sensors of all three types were fabricated, calibrated, and endurance tested. All three types of sensors met the fabricability survivability, and accuracy requirements established for their application.

  13. Comparison of the high temperature heat flux sensor to traditional heat flux gages under high heat flux conditions.

    SciTech Connect

    Blanchat, Thomas K.; Hanks, Charles R.

    2013-04-01

    Four types of heat flux gages (Gardon, Schmidt-Boelter, Directional Flame Temperature, and High Temperature Heat Flux Sensor) were assessed and compared under flux conditions ranging between 100-1000 kW/m2, such as those seen in hydrocarbon fire or propellant fire conditions. Short duration step and pulse boundary conditions were imposed using a six-panel cylindrical array of high-temperature tungsten lamps. Overall, agreement between all gages was acceptable for the pulse tests and also for the step tests. However, repeated tests with the HTHFS with relatively long durations at temperatures approaching 1000ÀC showed a substantial decrease (10-25%) in heat flux subsequent to the initial test, likely due to the mounting technique. New HTHFS gages have been ordered to allow additional tests to determine the cause of the flux reduction.

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

    SciTech Connect

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

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

  15. Ultra high temperature particle bed reactor design

    NASA Astrophysics Data System (ADS)

    Lazareth, Otto; Ludewig, Hans; Perkins, K.; Powell, J.

    1990-04-01

    A direct nuclear propulsion engine which could be used for a mission to Mars is designed. The main features of this reactor design are high values for I(sub sp) and very efficient cooling. This particle bed reactor consists of 37 cylindrical fuel elements embedded in a cylinder of beryllium which acts as a moderator and reflector. The fuel consists of a packed bed of spherical fissionable fuel particles. Gaseous H2 passes over the fuel bed, removes the heat, and is exhausted out of the rocket. The design was found to be neutronically critical and to have tolerable heating rates. Therefore, this particle bed reactor design is suitable as a propulsion unit for this mission.

  16. Advanced Models of LWR Pressure Vessel Embrittlement for Low Flux-HighFluence Conditions

    SciTech Connect

    Odette, G. Robert; Yamamoto, Takuya

    2013-06-17

    Neutron embrittlement of reactor pressure vessels (RPVs) is an unresolved issue for light water reactor life extension, especially since transition temperature shifts (TTS) must be predicted for high 80-year fluence levels up to approximately 1,020 n/cm{sup 2}, far beyond the current surveillance database. Unfortunately, TTS may accelerate at high fluence, and may be further amplified by the formation of late blooming phases that result in severe embrittlement even in low-copper (Cu) steels. Embrittlement by this mechanism is a potentially significant degradation phenomenon that is not predicted by current regulatory models. This project will focus on accurately predicting transition temperature shifts at high fluence using advanced physically based, empirically validated and calibrated models. A major challenge is to develop models that can adjust test reactor data to account for flux effects. Since transition temperature shifts depend on synergistic combinations of many variables, flux-effects cannot be treated in isolation. The best current models systematically and significantly under-predict transition temperature at high fluence, although predominantly for irradiations at much higher flux than actual RPV service. This project will integrate surveillance, test reactor and mechanism data with advanced models to address a number of outstanding RPV embrittlement issues. The effort will include developing new databases and preliminary models of flux effects for irradiation conditions ranging from very low (e.g., boiling water reactor) to high (e.g., accelerated test reactor). The team will also develop a database and physical models to help predict the conditions for the formation of Mn-Ni-Si late blooming phases and to guide future efforts to fully resolve this issue. Researchers will carry out other tasks on a best-effort basis, including prediction of transition temperature shift attenuation through the vessel wall, remediation of embrittlement by annealing

  17. Lessons Learned about Liquid Metal Reactors from FFTF Experience

    SciTech Connect

    Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.; Burke, Thomas M.; Grandy, Christopher

    2016-09-20

    The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens. In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports

  18. TEST-HOLE CONSTRUCTION FOR A NEUTRONIC REACTOR

    DOEpatents

    Ohlinger, L.A.; Seitz, F.; Young, G.J.

    1959-02-17

    Test-hole construction is described for a reactor which provides safe and ready access to the neutron flux region for specimen materials which are to be irradiated therein. An elongated tubular thimble adapted to be inserted in the access hole through the wall of the reactor is constructed of aluminum and is provided with a plurality of holes parallel to the axis of the thimble for conveying the test specimens into position for irradiation, and a conduit for the circulation of coolant. A laminated shield formed of alternate layers of steel and pressed wood fiber is disposed lengthwise of the thimble near the outer end thereof.

  19. Ion Engine Test Firing

    NASA Technical Reports Server (NTRS)

    1999-01-01

    This image of a xenon ion engine, photographed through a port of the vacuum chamber where it was being tested at NASA's Jet Propulsion Laboratory, shows the faint blue glow of charged atoms being emitted from the engine. The ion propulsion engine is the first non-chemical propulsion to be used as the primary means of propelling a spacecraft. The first flight in NASA's New Millennium Program, Deep Space 1 is designed to validate 12 new technologies for scientific space missions of the next century. Ion propulsion was first proposed in the 1950s and NASA performed experiments on this highly efficient propulsion system in the 1960s, but it was not used aboard an American spacecraft until the 1990s. Deep Space 1 was launched in October 1998 as part of NASA's New Millennium Program, which is managed by JPL for NASA's Office of Space Science, Washington, DC. The California Institute of Technology in Pasadena manages JPL for NASA. The almost imperceptible thrust from the ion propulsion system is equivalent to the pressure exerted by a sheet of paper held in the palm of your hand. The ion engine is very slow to pick up speed, but over the long haul it can deliver 10 times as much thrust per pound of fuel as more traditional rockets. Unlike the fireworks of most chemical rockets using solid or liquid fuels, the ion drive emits only an eerie blue glow as ionized (electrically charged) atoms of xenon are pushed out of the engine. Xenon is the same gas found in photo flash tubes and many lighthouse bulbs.

  20. TEST CELLS SE-5 - SE-8 - SE-10 IN THE ENGINE RESEARCH BUILDING ERB AND 117 HIGH ENERGY FUELS LABORAT

    NASA Technical Reports Server (NTRS)

    1963-01-01

    TEST CELLS SE-5 - SE-8 - SE-10 IN THE ENGINE RESEARCH BUILDING ERB AND 117 HIGH ENERGY FUELS LABORATORY HEFL - TRANSDUCER INSTRUMENTATION CONSOLE SE-10 - TEMPERATURE INSTRUMENTATION CONSOLE SE-10 - MODULE FUEL CELL EXPERIMENT SE-8 -

  1. TEST CELLS SE-5 - SE-8 - SE-10 IN THE ENGINE RESEARCH BUILDING ERB AND 117 HIGH ENERGY FUELS LABORAT

    NASA Technical Reports Server (NTRS)

    1963-01-01

    TEST CELLS SE-5 - SE-8 - SE-10 IN THE ENGINE RESEARCH BUILDING ERB AND 117 HIGH ENERGY FUELS LABORATORY HEFL - TRANSDUCER INSTRUMENTATION CONSOLE SE-10 - TEMPERATURE INSTRUMENTATION CONSOLE SE-10 - MODULE FUEL CELL EXPERIMENT SE-8

  2. TEST CELLS SE-5 - SE-8 - SE-10 IN THE ENGINE RESEARCH BUILDING ERB AND 117 HIGH ENERGY FUELS LABORAT

    NASA Technical Reports Server (NTRS)

    1963-01-01

    TEST CELLS SE-5 - SE-8 - SE-10 IN THE ENGINE RESEARCH BUILDING ERB AND 117 HIGH ENERGY FUELS LABORATORY HEFL - TRANSDUCER INSTRUMENTATION CONSOLE SE-10 - TEMPERATURE INSTRUMENTATION CONSOLE SE-10 - MODUEL FUEL CELL EXPERIMENT SE-8

  3. Compact steady-state and high-flux Falcon ion source for tests of plasma-facing materials

    SciTech Connect

    Girka, O.; Bizyukov, I.; Sereda, K.; Bizyukov, A.; Gutkin, M.; Sleptsov, V.

    2012-08-15

    This paper describes the design and operation of the Falcon ion source. It is based on conventional design of anode layer thrusters. This ion source is a versatile, compact, affordable, and highly functional in the research field of the fusion materials. The reversed magnetic field configuration of the source allows precise focusing of the ion beam into small spot of Almost-Equal-To 3 mm and also provides the limited capabilities for impurity mass-separation. As the result, the source generates steady-state ion beam, which irradiates surface with high heat (0.3 - 21 MW m{sup -2}) and particle fluxes (4 Multiplication-Sign 10{sup 21}- 3 Multiplication-Sign 10{sup 23} m{sup -2}s{sup -1}), which approaches the upper limit for the flux range expected in ITER.

  4. The Very High Temperature Reactor

    SciTech Connect

    Hans D. Gougar; David A. Petti

    2011-06-01

    The High Temperature Reactor (HTR) and Very High Temperature Reactor (VHTR) are types of nuclear power plants that, as the names imply, operate at temperatures above those of the conventional nuclear power plants that currently generate electricity in the US and other countries. Like existing nuclear plants, heat generated from the fission of uranium or plutonium atoms is carried off by a working fluid and can be used generate electricity. The very hot working fluid also enables the VHTR to drive other industrial processes that require high temperatures not achievable by conventional nuclear plants (Figure 1). For this reason, the VHTR is being considered for non-electrical energy applications. The reactor and power conversion system are constructed using special materials that make a core meltdown virtually impossible.

  5. Liquid Rocket Engine Testing

    NASA Technical Reports Server (NTRS)

    Rahman, Shamim

    2005-01-01

    Comprehensive Liquid Rocket Engine testing is essential to risk reduction for Space Flight. Test capability represents significant national investments in expertise and infrastructure. Historical experience underpins current test capabilities. Test facilities continually seek proactive alignment with national space development goals and objectives including government and commercial sectors.

  6. Advanced expander test bed engine

    NASA Technical Reports Server (NTRS)

    Mitchell, J. P.

    1992-01-01

    The Advanced Expander Test Bed (AETB) is a key element in NASA's Space Chemical Engine Technology Program for development and demonstration of expander cycle oxygen/hydrogen engine and advanced component technologies applicable to space engines as well as launch vehicle upper stage engines. The AETB will be used to validate the high pressure expander cycle concept, study system interactions, and conduct studies of advanced mission focused components and new health monitoring techniques in an engine system environment. The split expander cycle AETB will operate at combustion chamber pressures up to 1200 psia with propellant flow rates equivalent to 20,000 lbf vacuum thrust.

  7. High Temperature Ultrasonic Transducers for In-Service Inspection of Liquid Metal Fast Reactors

    SciTech Connect

    Griffin, Jeffrey W.; Posakony, Gerald J.; Harris, Robert V.; Baldwin, David L.; Jones, Anthony M.; Bond, Leonard J.

    2011-12-31

    In-service inspection of liquid metal (sodium) fast reactors requires the use of ultrasonic transducers capable of operating at high temperatures (>200°C), high gamma radiation fields, and the chemically reactive liquid sodium environment. In the early- to mid-1970s, the U.S. Atomic Energy Commission supported development of high-temperature, submersible single-element transducers, used for scanning and under-sodium imaging in the Fast Flux Test Facility and the Clinch River Breeder Reactor. Current work is building on this technology to develop the next generation of high-temperature linear ultrasonic transducer arrays for under-sodium viewing and in-service inspections.

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

    SciTech Connect

    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.

  9. Corrosion of spent Advanced Test Reactor fuel

    SciTech Connect

    Lundberg, L.B.; Croson, M.L.

    1994-11-01

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented.

  10. Preliminary neutronic analysis of a cavity test reactor

    NASA Technical Reports Server (NTRS)

    Whitmarsh, C. L., Jr.

    1973-01-01

    A reference configuration was calculated for a cavity test reactor to be used for testing the gascore nuclear rocket concept. A thermal flux of 4.1 x 10 to the 14th power neutrons per square centimeter per second in the cavity was provided by a driver fuel loading of 6.4 kg of enriched uranium in MTR fuel elements. The reactor was moderated and cooled by heavy water and reflected with 25.4 cm of beryllium. Power generation of 41.3 MW in the driver fuel is rejected to a heat sink. Design effort was directed toward minimization of driver power while maintaining 2.7 MW in the cavity during a test run. Ancillary data on material reactivity worths, reactivity coefficients, flux spectra, and power distributions are reported.

  11. Testing of high-octane fuels in the single-cylinder airplane engine

    NASA Technical Reports Server (NTRS)

    Seeber, Fritz

    1940-01-01

    One of the most important properties of aviation fuels for spark-ignition engines is their knock rating. The CFR engine tests of fuels of 87 octane and above does not always correspond entirely to the actual behavior of these fuels in the airplane engine. A method is therefore developed which, in contrast to the octane number determination, permits a testing of the fuel under various temperatures and fuel mixture conditions. The following reference fuels were employed: 1) Primary fuels; isooctane and n-heptane; 2) Secondary fuels; pure benzene and synthetic benzine.

  12. Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

    NASA Technical Reports Server (NTRS)

    Tong, H.; Snow, G. C.; Chu, E. K.; Chang, R. L. S.; Angwin, M. J.; Pessagno, S. L.

    1981-01-01

    Durable catalytic reactors for advanced gas turbine engines were developed. Objectives were: to evaluate furnace aging as a cost effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1,000 hours of combustion durability, and define a catalytic reactor system with a high probability of successful integration into an automotive gas turbine engine. Fourteen different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel for combustion conditions. Eight additional catalytic reactors were evaluated and one of these was successfully combustion tested on propane fuel. This durability reactor used graded cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

  13. Space shuttle maneuvering engine reusable thrust chamber program. Task 11: High epsilon injector test report

    NASA Technical Reports Server (NTRS)

    Pauckert, R. P.

    1974-01-01

    A subscale injector test program conducted with a ten-inch diameter like-doublet injector to determine performance and heat flux profiles vs chamber pressure and propellant mixture ratio, and to determine stability characteristics with various acoustic cavity configurations is described. Results are presented.

  14. Critical heat flux test apparatus

    DOEpatents

    Welsh, Robert E.; Doman, Marvin J.; Wilson, Edward C.

    1992-01-01

    An apparatus for testing, in situ, highly irradiated specimens at high temperature transients is provided. A specimen, which has a thermocouple device attached thereto, is manipulated into test position in a sealed quartz heating tube by a robot. An induction coil around a heating portion of the tube is powered by a radio frequency generator to heat the specimen. Sensors are connected to monitor the temperatures of the specimen and the induction coil. A quench chamber is located below the heating portion to permit rapid cooling of the specimen which is moved into this quench chamber once it is heated to a critical temperature. A vacuum pump is connected to the apparatus to collect any released fission gases which are analyzed at a remote location.

  15. Development of a 10-decade single-mode reactor flux monitoring system

    SciTech Connect

    Valentine, K.H.; Shepard, R.L.; Falter, K.G.; Reese, W.B.

    1988-03-31

    Conventional wide-range neutron channels employ three optional modes to monitor the required flux range from source levels to full power (typically 10 or more decades). Difficult calibrations are necessary to provide a continuous output signal when such a system switches from counting mode in the source range to mean-square voltage mode in the midrange to dc current mode in the power range. In an ORNL proof-of-principle test, a method of extended range counting was implemented with a fission counter and conventional wide-band pulse processing electronics to provide a single-mode, monotonically increasing signal that spanned /approximately 10/ decades of neutron flux. Ongoing work includes design, fabrication, and testing of a comlpete neutron flux monitoring system suitable for advanced liquid metal reactor designs. 6 refs., 4 figs.

  16. High temperature catalytic membrane reactors

    SciTech Connect

    Not Available

    1990-03-01

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

  17. Studies Related to the Oregon State University High Temperature Test Facility: Scaling, the Validation Matrix, and Similarities to the Modular High Temperature Gas-Cooled Reactor

    SciTech Connect

    Richard R. Schultz; Paul D. Bayless; Richard W. Johnson; William T. Taitano; James R. Wolf; Glenn E. McCreery

    2010-09-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5 year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant project. Because the NRC interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC). Since DOE has incorporated the HTTF as an ingredient in the NGNP thermal-fluids validation program, several important outcomes should be noted: 1. The reference prismatic reactor design, that serves as the basis for scaling the HTTF, became the modular high temperature gas-cooled reactor (MHTGR). The MHTGR has also been chosen as the reference design for all of the other NGNP thermal-fluid experiments. 2. The NGNP validation matrix is being planned using the same scaling strategy that has been implemented to design the HTTF, i.e., the hierarchical two-tiered scaling methodology developed by Zuber in 1991. Using this approach a preliminary validation matrix has been designed that integrates the HTTF experiments with the other experiments planned for the NGNP thermal-fluids verification and validation project. 3. Initial analyses showed that the inherent power capability of the OSU infrastructure, which only allowed a total operational facility power capability of 0.6 MW, is

  18. The Advanced Test Reactor Irradiation Capabilities Available as a National Scientific User Facility

    SciTech Connect

    S. Blaine Grover

    2008-09-01

    The Advanced Test Reactor (ATR) is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These capabilities include simple capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. Monitoring systems have also been utilized to monitor different parameters such as fission gases for fuel experiments, to measure specimen performance during irradiation. ATR’s control system provides a stable axial flux profile throughout each reactor operating cycle, and allows the thermal and fast neutron fluxes to be controlled separately in different sections of the core. The ATR irradiation positions vary in diameter from 16 mm to 127 mm over an active core height of 1.2 m. This paper discusses the different irradiation capabilities with examples of different experiments and the cost/benefit issues related to each capability. The recent designation of ATR as a national scientific user facility will make the ATR much more accessible at very low to no cost for research by universities and possibly commercial entities.

  19. High-temperature high-bandwidth fiber optic MEMS pressure-sensor technology for turbine engine component testing

    NASA Astrophysics Data System (ADS)

    Pulliam, Wade J.; Russler, Patrick M.; Fielder, Robert S.

    2002-02-01

    Acquiring accurate, transient measurements in harsh environments has always pushed the limits of available measurement technology. Until recently, the technology to directly measure certain properties in extremely high temperature environments has not existed. Advancements in optical measurement technology have led to the development of measurement techniques for pressure, temperature, acceleration, skin friction, etc. using extrinsic Fabry-Perot interferometry (EFPI). The basic operating principle behind EFPI enables the development of sensors that can operate in the harsh conditions associated with turbine engines, high-speed combustors, and other aerospace propulsion applications where the flow environment is dominated by high frequency pressure and temperature variations caused by combustion instabilities, blade-row interactions, and unsteady aerodynamic phenomena. Using micromachining technology, these sensors are quite small and therefore ideal for applications where restricted space or minimal measurement interference is a consideration. In order to help demonstrate the general functionality of this measurement technology, sensors and signal processing electronics currently under development by Luna Innovations were used to acquire point measurements during testing of a transonic fan in the Compressor Research Facility (CRF) at the Turbine Engine Research Center (TERC), WPAFB. Acquiring pressure measurements at the surface of the casing wall provides data that are useful in understanding the effects of pressure fluctuations on the operation and lifetime wear of a fan. This measurement technique is useful in both test rig applications and in operating engines where lifetime wear characterization is important. The measurements acquired during this test also assisted in the continuing development of this technology for higher temperature environments by providing proof-of-concept data for sensors based on advanced microfabrication and optical techniques.

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

    SciTech Connect

    Agarwal, Vivek; Smith, James A.; Jewell, James Keith

    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.

  1. Crash tests of four identical high-wing single-engine airplanes

    NASA Technical Reports Server (NTRS)

    Vaughan, V. L., Jr.; Hayduk, R. J.

    1980-01-01

    Four identical four place, high wing, single engine airplane specimens with nominal masses of 1043 kg were crash tested at the Langley Impact Dynamics Research Facility under controlled free flight conditions. These tests were conducted with nominal velocities of 25 m/sec along the flight path angles, ground contact pitch angles, and roll angles. Three of the airplane specimens were crashed on a concrete surface; one was crashed on soil. Crash tests revealed that on a hard landing, the main landing gear absorbed about twice the energy for which the gear was designed but sprang back, tending to tip the airplane up to its nose. On concrete surfaces, the airplane impacted and remained in the impact attitude. On soil, the airplane flipped over on its back. The crash impact on the nose of the airplane, whether on soil or concrete, caused massive structural crushing of the forward fuselage. The liveable volume was maintained in both the hard landing and the nose down specimens but was not maintained in the roll impact and nose down on soil specimens.

  2. High flux solar energy transformation

    DOEpatents

    Winston, Roland; Gleckman, Philip L.; O'Gallagher, Joseph J.

    1991-04-09

    Disclosed are multi-stage systems for high flux transformation of solar energy allowing for uniform solar intensification by a factor of 60,000 suns or more. Preferred systems employ a focusing mirror as a primary concentrative device and a non-imaging concentrator as a secondary concentrative device with concentrative capacities of primary and secondary stages selected to provide for net solar flux intensification of greater than 2000 over 95 percent of the concentration area. Systems of the invention are readily applied as energy sources for laser pumping and in other photothermal energy utilization processes.

  3. High flux solar energy transformation

    DOEpatents

    Winston, R.; Gleckman, P.L.; O'Gallagher, J.J.

    1991-04-09

    Disclosed are multi-stage systems for high flux transformation of solar energy allowing for uniform solar intensification by a factor of 60,000 suns or more. Preferred systems employ a focusing mirror as a primary concentrative device and a non-imaging concentrator as a secondary concentrative device with concentrative capacities of primary and secondary stages selected to provide for net solar flux intensification of greater than 2000 over 95 percent of the concentration area. Systems of the invention are readily applied as energy sources for laser pumping and in other photothermal energy utilization processes. 7 figures.

  4. Trajectory Optimization and Conceptual Study of Small Test Vehicles for Hypersonic Engine Using High-Altitude Balloon

    NASA Astrophysics Data System (ADS)

    Tsuchiya, Takeshi; Takenaka, Youichi; Taguchi, Hideyuki; Sawai, Shujiro

    Japan Aerospace Exploration Agency, JAXA announced a long-term vision recently. In the vision, JAXA aims to develop hypersonic aircrafts. A pre-cooled turbojet engine has great potential as one of newly developed hypersonic air-breathing engines. We also expect the engine to be installed in space transportation vehicles in future. For combustion test in real flight condition of the engines, JAXA has an experimental plan with a small test vehicle falling from a high-altitude balloon. This paper applies numerical analysis and optimization techniques to conceptual designs of the test vehicle in order to obtain the best configuration and trajectory that can achieve the flight test. The results show helpful knowledge when we design prototype vehicles.

  5. Trajectory Optimization and Conceptual Study of Small Test Vehicles for a Hypersonic Engine Using a High-Altitude Balloon

    NASA Astrophysics Data System (ADS)

    Tsuchiya, Takeshi; Takenaka, Youichi; Taguchi, Hideyuki; Sawai, Shujiro

    The Japan Aerospace Exploration Agency, JAXA, announced a long-term vision recently. In the vision, JAXA aims to develop hypersonic aircrafts. A pre-cooled turbojet engine has great potential as one of newly developed hypersonic airbreathing engines. We also expect the engine to be installed in space transportation vehicles in the future. For combustion test in the real flight conditions of the engines, JAXA has an experimental plan where a small test vehicle is released from a high-altitude balloon. This paper applies numerical analysis and optimization techniques to conceptual designs of the test vehicle in order to obtain the best configuration and trajectory for the flight test. The results show helpful knowledge for designing prototype vehicles.

  6. J-2 Engine Test

    NASA Technical Reports Server (NTRS)

    1963-01-01

    Smokeless flame juts from the diffuser of a unique vacuum chamber in which the upper stage rocket engine, the hydrogen fueled J-2, was tested at a simulated space altitude in excess of 60,000 feet. The smoke you see is actually steam. In operation, vacuum is established by injecting steam into the chamber and is maintained by the thrust of the engine firing through the diffuser. The engine was tested in this environment for start, stop, coast, restart, and full-duration operations. The chamber was located at Rocketdyne's Propulsion Field Laboratory, in the Santa Susana Mountains, near Canoga Park, California. The J-2 engine was developed by Rocketdyne for the Marshall Space Flight Center.

  7. Solar-Thermal Engine Testing

    NASA Technical Reports Server (NTRS)

    Tucker, Stephen; Salvail, Pat; Haynes, Davy (Technical Monitor)

    2001-01-01

    A solar-thermal engine serves as a high-temperature solar-radiation absorber, heat exchanger, and rocket nozzle. collecting concentrated solar radiation into an absorber cavity and transferring this energy to a propellant as heat. Propellant gas can be heated to temperatures approaching 4,500 F and expanded in a rocket nozzle, creating low thrust with a high specific impulse (I(sub sp)). The Shooting Star Experiment (SSE) solar-thermal engine is made of 100 percent chemical vapor deposited (CVD) rhenium. The engine 'module' consists of an engine assembly, propellant feedline, engine support structure, thermal insulation, and instrumentation. Engine thermal performance tests consist of a series of high-temperature thermal cycles intended to characterize the propulsive performance of the engines and the thermal effectiveness of the engine support structure and insulation system. A silicone-carbide electrical resistance heater, placed inside the inner shell, substitutes for solar radiation and heats the engine. Although the preferred propellant is hydrogen, the propellant used in these tests is gaseous nitrogen. Because rhenium oxidizes at elevated temperatures, the tests are performed in a vacuum chamber. Test data will include transient and steady state temperatures on selected engine surfaces, propellant pressures and flow rates, and engine thrust levels. The engine propellant-feed system is designed to Supply GN2 to the engine at a constant inlet pressure of 60 psia, producing a near-constant thrust of 1.0 lb. Gaseous hydrogen will be used in subsequent tests. The propellant flow rate decreases with increasing propellant temperature, while maintaining constant thrust, increasing engine I(sub sp). In conjunction with analytical models of the heat exchanger, the temperature data will provide insight into the effectiveness of the insulation system, the structural support system, and the overall engine performance. These tests also provide experience on operational

  8. Engineering problems of tandem-mirror reactors

    SciTech Connect

    Moir, R.W.; Barr, W.L.; Boghosian, B.M.

    1981-10-22

    We have completed a comparative evaluation of several end plug configurations for tandem mirror fusion reactors with thermal barriers. The axi-cell configuration has been selected for further study and will be the basis for a detailed conceptual design study to be carried out over the next two years. The axi-cell end plug has a simple mirror cell produced by two circular coils followed by a transition coil and a yin-yang pair, which provides for MHD stability. This paper discusses some of the many engineering problems facing the designer. We estimated the direct cost to be 2$/W/sub e/. Assuming total (direct and indirect) costs to be twice this number, we need to reduce total costs by factors between 1.7 and 2.3 to compete with future LWRs levelized cost of electricity. These reductions may be possible by designing magnets producing over 20T made possible by use of combinations of superconducting and normal conducting coils as well as improvements in performance and cost of neutral beam and microwave power systems. Scientific and technological understanding and innovation are needed in the area of thermal barrier pumping - a process by which unwanted particles are removed (pumped) from certain regions of velocity and real space in the end plug. Removal of exhaust fuel ions, fusion ash and impurities by action of a halo plasma and plasma dump in the mirror end region is another challenging engineering problem discussed in this paper.

  9. Enhanced In-pile Instrumentation for Material Testing Reactors

    SciTech Connect

    Joy Rempe; Darrell Knudson; Joshua Daw; Troy Unruh; Benjamin Chase; Kurt Davis; Robert Schley

    2012-07-01

    An increasing number of U.S. nuclear research programs are requesting enhanced in-pile instrumentation capable of providing real-time measurements of key parameters during irradiations. For example, fuel research and development funded by the U.S. Department of Energy now emphasize approaches that rely on first principle models to develop optimized fuel designs that offer significant improvements over current fuels. To facilitate this approach, high fidelity, real-time data are essential for characterizing the performance of new fuels during irradiation testing. Furthermore, sensors that obtain such data must be miniature, reliable and able to withstand high flux/high temperature conditions. Depending on user requirements, sensors may need to obtain data in inert gas, pressurized water, or liquid metal environments. To address these user needs, in-pile instrumentation development efforts have been initiated as part of the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF), the Fuel Cycle Research & Development (FCR&D), and the Nuclear Energy Enabling Technology (NEET) programs. This paper reports on recent INL achievements to support these programs. Specifically, an overview of the types of sensors currently available to support in-pile irradiations and those sensors currently available to MTR users are identified. In addition, recent results and products available from sensor research and development are detailed. Specifically, progress in deploying enhanced in-pile sensors for detecting elongation and thermal conductivity are reported. Results from research to evaluate the viability of ultrasonic and fiber optic technologies for irradiation testing are also summarized.

  10. Modeling of Highly Instrumented Honeywell Turbofan Engine Tested with Ice Crystal Ingestion in the NASA Propulsion System Laboratory

    NASA Technical Reports Server (NTRS)

    Veres, Joseph P.; Jorgenson, Philip C. E.; Jones, Scott M.

    2016-01-01

    The Propulsion Systems Laboratory (PSL), an altitude test facility at NASA Glenn Research Center, has been used to test a highly instrumented turbine engine at simulated altitude operating conditions. This is a continuation of the PSL testing that successfully duplicated the icing events that were experienced in a previous engine (serial LF01) during flight through ice crystal clouds, which was the first turbofan engine tested in PSL. This second model of the ALF502R-5A serial number LF11 is a highly instrumented version of the previous engine. The PSL facility provides a continuous cloud of ice crystals with controlled characteristics of size and concentration, which are ingested by the engine during operation at simulated altitudes. Several of the previous operating points tested in the LF01 engine were duplicated to confirm repeatability in LF11. The instrumentation included video cameras to visually illustrate the accretion of ice in the low pressure compressor (LPC) exit guide vane region in order to confirm the ice accretion, which was suspected during the testing of the LF01. Traditional instrumentation included static pressure taps in the low pressure compressor inner and outer flow path walls, as well as total pressure and temperature rakes in the low pressure compressor region. The test data was utilized to determine the losses and blockages due to accretion in the exit guide vane region of the LPC. Multiple data points were analyzed with the Honeywell Customer Deck. A full engine roll back point was modeled with the Numerical Propulsion System Simulation (NPSS) code. The mean line compressor flow analysis code with ice crystal modeling was utilized to estimate the parameters that indicate the risk of accretion, as well as to estimate the degree of blockage and losses caused by accretion during a full engine roll back point. The analysis provided additional validation of the icing risk parameters within the LPC, as well as the creation of models for

  11. Diesel Engine Idling Test

    SciTech Connect

    Larry Zirker; James Francfort; Jordon Fielding

    2006-02-01

    In support of the Department of Energy’s FreedomCAR and Vehicle Technology Program Office goal to minimize diesel engine idling and reduce the consumption of millions of gallons of diesel fuel consumed during heavy vehicle idling periods, the Idaho National Laboratory (INL) conducted tests to characterize diesel engine wear rates caused by extended periods of idling. INL idled two fleet buses equipped with Detroit Diesel Series 50 engines, each for 1,000 hours. Engine wear metals were characterized from weekly oil analysis samples and destructive filter analyses. Full-flow and the bypass filter cartridges were removed at four stages of the testing and sent to an oil analysis laboratory for destructive analysis to ascertain the metals captured in the filters and to establish wear rate trends. Weekly samples were sent to two independent oil analysis laboratories. Concurrent with the filter analysis, a comprehensive array of other laboratory tests ascertained the condition of the oil, wear particle types, and ferrous particles. Extensive ferrogram testing physically showed the concentration of iron particles and associated debris in the oil. The tests results did not show the dramatic results anticipated but did show wear trends. New West Technologies, LLC, a DOE support company, supplied technical support and data analysis throughout the idle test.

  12. Heat-flux gage measurements on a flat plate at a Mach number of 4.6 in the VSD high speed wind tunnel, a feasibility test (LA28). [wind tunnel tests of measuring instruments for boundary layer flow

    NASA Technical Reports Server (NTRS)

    1975-01-01

    The feasibility of employing thin-film heat-flux gages was studied as a method of defining boundary layer characteristics at supersonic speeds in a high speed blowdown wind tunnel. Flow visualization techniques (using oil) were employed. Tabulated data (computer printouts), a test facility description, and photographs of test equipment are given.

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

    PubMed

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

    2015-10-01

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

  14. Proceedings of the Oak Ridge National Laboratory/Brookhaven National Laboratory workshop on neutron scattering instrumentation at high-flux reactors

    SciTech Connect

    McBee, M.R.; Axe, J.D.; Hayter, J.B.

    1990-07-01

    For the first three decades following World War II, the US, which pioneered the field of neutron scattering research, enjoyed uncontested leadership in the field. By the mid-1970's, other countries, most notably through the West European consortium at Institut Laue-Langevin (ILL) in Grenoble, France, had begun funding neutron scattering on a scale unmatched in this country. By the early 1980's, observers charged with defining US scientific priorities began to stress the need for upgrading and expansion of US research reactor facilities. The conceptual design of the ANS facility is now well under way, and line-item funding for more advanced design is being sought for FY 1992. This should lead to a construction request in FY 1994 and start-up in FY 1999, assuming an optimal funding profile. While it may be too early to finalize designs for instruments whose construction is nearly a decade removed, it is imperative that we begin to develop the necessary concepts to ensure state-of-the-art instrumentation for the ANS. It is in this context that this Instrumentation Workshop was planned. The workshop touched upon many ideas that must be considered for the ANS, and as anticipated, several of the discussions and findings were relevant to the planning of the HFBR Upgrade. In addition, this report recognizes numerous opportunities for further breakthroughs on neutron instrumentation in areas such as improved detection schemes (including better tailored scintillation materials and image plates, and increased speed in both detection and data handling), in-beam monitors, transmission white beam polarizers, multilayers and supermirrors, and more. Each individual report has been cataloged separately.

  15. Poloidal flux linkage requirements for the International Thermonuclear Experimental Reactor

    SciTech Connect

    Jardin, S.C.; Kessel, C.; Pomphrey, N.

    1994-01-01

    We have applied two computational models to calculate the poloidal flux linkage requirements for the current ramp-up and for the flattop phase of the proposed International Thermonuclear Experimental Reactor (ITER). For the current ramp-up phase, we have used the TSC code to simulate the entire current ramp-up period as described in the TAC-3 Physics Report. We have extended the time of the simulation to cover the full current penetration time, that is, until the loop voltage is a constant throughout the plasma. Sensitivity studies have been performed with respect to current ramp-up time, impurity concentration, and to the time of onset of auxiliary heating. We have also used a steady state plasma equilibrium code that has the constant loop voltage constraint built in to survey the dependence of the steady state loop-voltage on the density and temperature profiles. This calculation takes into account the plasma bootstrap current contribution, including non-circular and collisional corrections. The results can be displayed as contours of the loop-voltage on a POPCON like diagram.

  16. Steel slag carbonation in a flow-through reactor system: the role of fluid-flux.

    PubMed

    Berryman, Eleanor J; Williams-Jones, Anthony E; Migdisov, Artashes A

    2015-01-01

    Steel production is currently the largest industrial source of atmospheric CO2. As annual steel production continues to grow, the need for effective methods of reducing its carbon footprint increases correspondingly. The carbonation of the calcium-bearing phases in steel slag generated during basic oxygen furnace (BOF) steel production, in particular its major constituent, larnite {Ca2SiO4}, which is a structural analogue of olivine {(MgFe)2SiO4}, the main mineral subjected to natural carbonation in peridotites, offers the potential to offset some of these emissions. However, the controls on the nature and efficiency of steel slag carbonation are yet to be completely understood. Experiments were conducted exposing steel slag grains to a CO2-H2O mixture in both batch and flow-through reactors to investigate the impact of temperature, fluid flux, and reaction gradient on the dissolution and carbonation of steel slag. The results of these experiments show that dissolution and carbonation of BOF steel slag are more efficient in a flow-through reactor than in the batch reactors used in most previous studies. Moreover, they show that fluid flux needs to be optimized in addition to grain size, pressure, and temperature, in order to maximize the efficiency of carbonation. Based on these results, a two-stage reactor consisting of a high and a low fluid-flux chamber is proposed for CO2 sequestration by steel slag carbonation, allowing dissolution of the slag and precipitation of calcium carbonate to occur within a single flow-through system.

  17. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOEpatents

    Lasche, G.P.

    1983-09-29

    The invention is a laser or particle-beam-driven fusion reactor system which takes maximum advantage of both the very short pulsed nature of the energy release of inertial confinement fusion (ICF) and the very small volumes within which the thermonuclear burn takes place. The pulsed nature of ICF permits dynamic direct energy conversion schemes such as magnetohydrodynamic (MHD) generation and magnetic flux compression; the small volumes permit very compact blanket geometries. By fully exploiting these characteristics of ICF, it is possible to design a fusion reactor with exceptionally high power density, high net electric efficiency, and low neutron-induced radioactivity. The invention includes a compact blanket design and method and apparatus for obtaining energy utilizing the compact blanket.

  18. Knowledge Management at the Fast Flux Test Facility

    SciTech Connect

    Wootan, David W.; Omberg, Ronald P.

    2013-06-01

    One of the goals of the Department of Energy’s Office of Nuclear Energy, initiated under the Fuel Cycle Research and Development Program (FCRD) and continued under the Advanced Reactor Concepts Program (ARC) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMRs) that could support the development of an environmentally and economically sound nuclear fuel cycle. The Fast Flux Test Facility (FFTF) is the most recent LMR to operate in the United States, from 1982 to 1992, and was designed as a fully instrumented test reactor with on-line, real time test control and performance monitoring of components and tests installed in the reactor. The 10 years of operation of the FFTF provided a very useful framework for testing the advances in LMR safety technology based on passive safety features that may be of increased importance to new designs after the events at Fukushima. Knowledge preservation at the FFTF is focused on the areas of design, construction, and startup of the reactor, as well as on preserving information obtained from 10 years of successful operating history and extensive irradiation testing of fuels and materials. In order to ensure protection of information at risk, the program to date has sequestered reports, files, tapes, and drawings to allow for secure retrieval. The FFTF knowledge management program includes a disciplined and orderly approach to respond to client’s requests for documents and data in order to minimize the search effort and ensure that future requests for this information can be readily accommodated.

  19. Completion of the first NGNP Advanced Gas Reactor Fuel Irradiation Experiment, AGR-1, in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover; John Maki; David Petti

    2010-10-01

    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). 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 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 completed a very successful irradiation in early November 2009. The design of AGR-1 test train and support systems used to monitor and control the experiment during

  20. Engineering Development of Slurry Bubble Column Reactor (SBCR) Technology

    SciTech Connect

    Puneet Gupta

    2002-07-31

    This report summarizes the procedures used and results obtained in determining radial gas holdup profiles, via gamma ray scanning, and in assessing liquid and gas mixing parameters, via radioactive liquid and gas tracers, during Fischer Tropsch synthesis. The objectives of the study were (i) to develop a procedure for detection of gas holdup radial profiles in operating reactors and (ii) to test the ability of the developed, previously described, engineering models to predict the observed liquid and gas mixing patterns. It was shown that the current scanning procedures were not precise enough to obtain an accurate estimate of the gas radial holdup profile and an improved protocol for future use was developed. The previously developed physically based model for liquid mixing was adapted to account for liquid withdrawal from the mid section of the column. The ability of our engineering mixing models for liquid and gas phase to predict both liquid and gas phase tracer response was established and illustrated.

  1. Structures for handling high heat fluxes

    NASA Astrophysics Data System (ADS)

    Watson, R. D.

    1990-12-01

    The divertor is reconized as one of the main performance limiting components for ITER. This paper reviews the critical issues for structures that are designed to withstand heat fluxes > 5 MW/m 2. High velocity, sub-cooled water with twisted tape inserts for enhanced heat transfer provides a critical heat flux limit of 40-60 MW/m 2. Uncertainties in physics and engineering heat flux peaking factors require that the design heat flux not exceed 10 MW/m 2 to maintain an adequate burnout safety margin. Armor tiles and heat sink materials must have a well matched thermal expansion coefficient to minimize stresses. The divertor lifetime from sputtering erosion is highly uncertain. The number of disruptions specified for ITER must be reduced to achieve a credible design. In-situ plasma spray repair with thick metallic coatings may reduce the problems of erosion. Runaway electrons in ITER have the potential to melt actively cooled components in a single event. A water leak is a serious accident because of steam reactions with hot carbon, beryllium, or tungsten that can mobilize large amounts of tritium and radioactive elements. If the plasma does not shutdown immediately, the divertor can melt in 1-10 s after a loss of coolant accident. Very high reliability of carbon tile braze joints will be required to achieve adequate safety and performance goals. Most of these critical issues will be addressed in the near future by operation of the Tore Supra pump limiters and the JET pumped divertor. An accurate understanding of the power flow out of edge of a DT burning plasma is essential to successful design of high heat flux components.

  2. Viscosity test standards for engine oils

    SciTech Connect

    Not Available

    1990-01-01

    This report presents a compilation of 10 ASTM standards that cover both low and high temperature viscosity tests for automotive engine oils, with respect to low temperature flow properties and performance requirements under high temperature, high shear rate conditions. Society of Automotive Engineer's Engine Oil Viscosity Classification SAE J300 is included to provide low temperature high shear rate method.

  3. Topical report: Natural convection shutdown heat removal test facility (NSTF) evaluation for generating additional reactor cavity cooling system (RCCS) data.

    SciTech Connect

    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

  4. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    SciTech Connect

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

  5. High temperature NASP engine seal development

    NASA Technical Reports Server (NTRS)

    Steinetz, Bruce M.; Melis, Matthew E.; Orletski, Dirk; Test, Mark G.

    1991-01-01

    Key to the development of advanced hypersonic engines such as those being considered for the National Aerospace Plane (NASP) is the development and evaluation of high temperature, flexible seals that must seal the many feet of gaps between the articulating and stationary engine panels. Recent seal progress made at NASA-Lewis is reviewed in the areas of seal concept maturation, test rig development, and performance tests. A test fixture was built at NASA capable of subjecting candidate 3 ft long seals to engine simulated temperatures (up to 1500 F), pressures (up to 100 psi), and engine wall distortions (up to 0.15 in only 18 in span). Leakage performance test results at high temperatures are presented for an innovative high temperature, flexible ceramic wafer seal. Also described is a joint Pratt and Whitney/NASA planned test program to evaluate thermal performance of a braided rope seal under engine simulated heat flux rates (up to 400 Btu/sq ft s), and supersonic flow conditions. These conditions are produced by subjecting the seal specimen to hydrogen oxygen rocket exhaust that flows tangent to the specimen.

  6. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.

    1963-06-11

    A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

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

    NASA Technical Reports Server (NTRS)

    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.

  8. Liquid Rocket Engine Testing Overview

    NASA Technical Reports Server (NTRS)

    Rahman, Shamim

    2005-01-01

    Contents include the following: Objectives and motivation for testing. Technology, Research and Development Test and Evaluation (RDT&E), evolutionary. Representative Liquid Rocket Engine (LRE) test compaigns. Apollo, shuttle, Expandable Launch Vehicles (ELV) propulsion. Overview of test facilities for liquid rocket engines. Boost, upper stage (sea-level and altitude). Statistics (historical) of Liquid Rocket Engine Testing. LOX/LH, LOX/RP, other development. Test project enablers: engineering tools, operations, processes, infrastructure.

  9. Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    SciTech Connect

    Dawn Scates

    2010-10-01

    A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

  10. On-line calibration of high-response pressure transducers during jet-engine testing

    NASA Technical Reports Server (NTRS)

    Armentrout, E. C.

    1974-01-01

    Jet engine testing is reported concerned with the effect of inlet pressure and temperature distortions on engine performance and involves the use of numerous miniature pressure transducers. Despite recent improvements in the manufacture of miniature pressure transducers, they still exhibit sensitivity change and zero-shift with temperature and time. To obtain meaningful data, a calibration system is needed to determine these changes. A system has been developed which provides for computer selection of appropriate reference pressures selected from nine different sources to provide a two- or three-point calibration. Calibrations are made on command, before and sometimes after each data point. A unique no leak matrix valve design is used in the reference pressure system. Zero-shift corrections are measured and the values are automatically inserted into the data reduction program.

  11. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    SciTech Connect

    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 from normal high

  12. Pulse Detonation Engine Test Bed Developed

    NASA Technical Reports Server (NTRS)

    Breisacher, Kevin J.

    2002-01-01

    A detonation is a supersonic combustion wave. A Pulse Detonation Engine (PDE) repetitively creates a series of detonation waves to take advantage of rapid burning and high peak pressures to efficiently produce thrust. NASA Glenn Research Center's Combustion Branch has developed a PDE test bed that can reproduce the operating conditions that might be encountered in an actual engine. It allows the rapid and cost-efficient evaluation of the technical issues and technologies associated with these engines. The test bed is modular in design. It consists of various length sections of both 2- and 2.6- in. internal-diameter combustor tubes. These tubes can be bolted together to create a variety of combustor configurations. A series of bosses allow instrumentation to be inserted on the tubes. Dynamic pressure sensors and heat flux gauges have been used to characterize the performance of the test bed. The PDE test bed is designed to utilize an existing calorimeter (for heat load measurement) and windowed (for optical access) combustor sections. It uses hydrogen as the fuel, and oxygen and nitrogen are mixed to simulate air. An electronic controller is used to open the hydrogen and air valves (or a continuous flow of air is used) and to fire the spark at the appropriate times. Scheduled tests on the test bed include an evaluation of the pumping ability of the train of detonation waves for use in an ejector and an evaluation of the pollutants formed in a PDE combustor. Glenn's Combustion Branch uses the National Combustor Code (NCC) to perform numerical analyses of PDE's as well as to evaluate alternative detonative combustion devices. Pulse Detonation Engine testbed.

  13. Temperature controlled material irradiation in the advanced test reactor

    SciTech Connect

    Furstenau, R.V.; Ingrahm, F.W.

    1995-12-31

    The Advanced Test Reactor (ATR) is located at the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho, USA and is owned and regulated by the U.S. Department of Energy (US DOE). The ATR is operated for the US DOE by Lockheed Martin Idaho Technologies. In recent years, prime irradiation space in the ATR has been made available for use by customers having irradiation service needs in addition to the reactor`s principal user, the U.S. Naval Nuclear Propulsion Program. To enhance the reactor`s capabilities, the US DOE has initiated the development of an Irradiation Test Vehicle (ITV) capable of providing neutron spectral tailoring and temperature control for up to 28 experiments. The ATR-ITV will have the flexibility to simultaneously support a variety of experiments requiring fast, thermal or mixed spectrum neutron environments. Temperature control is accomplished by varying the thermal conductivity across a gas gap established between the experiment specimen capsule wall and the experiment `in-pile tube (IPT)` inside diameter. Thermal conductivity is adjusted by alternating the control gas mixture ratio of two gases with different thermal conductivities.

  14. Fast Flux Test Facility core restraint system performance

    SciTech Connect

    Hecht, S.L.; Trenchard, R.G.

    1990-02-01

    Characterizing Fast Flux Test Facility (FFTF) core restraint system performance has been ongoing since the first operating cycle. Characterization consists of prerun analysis for each core load, in-reactor and postirradiation measurements of subassembly withdrawal loads and deformations, and using measurement data to fine tune predictive models. Monitoring FFTF operations and performing trend analysis has made it possible to gain insight into core restraint system performance and head off refueling difficulties while maximizing component lifetimes. Additionally, valuable information for improved designs and operating methods has been obtained. Focus is on past operating experience, emphasizing performance improvements and avoidance of potential problems. 4 refs., 12 figs., 2 tabs.

  15. Microstructural Characterization of a Mg Matrix U-Mo Dispersion Fuel Plate Irradiated in the Advanced Test Reactor to High Fission Density: SEM Results

    NASA Astrophysics Data System (ADS)

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adam B.; Medvedev, Pavel G.; Madden, James W.; Moore, Glenn A.

    2016-06-01

    Low-enriched (U-235 <20 pct) U-Mo dispersion fuel is being developed for use in research and test reactors. In most cases, fuel plates with Al or Al-Si alloy matrices have been tested in the Advanced Test Reactor to support this development. In addition, fuel plates with Mg as the matrix have also been tested. The benefit of using Mg as the matrix is that it potentially will not chemically interact with the U-Mo fuel particles during fabrication or irradiation, whereas with Al and Al-Si alloys such interactions will occur. Fuel plate R9R010 is a Mg matrix fuel plate that was aggressively irradiated in ATR. This fuel plate was irradiated as part of the RERTR-8 experiment at high temperature, high fission rate, and high power, up to high fission density. This paper describes the results of the scanning electron microscopy (SEM) analysis of an irradiated fuel plate using polished samples and those produced with a focused ion beam. A follow-up paper will discuss the results of transmission electron microscopy (TEM) analysis. Using SEM, it was observed that even at very aggressive irradiation conditions, negligible chemical interaction occurred between the irradiated U-7Mo fuel particles and Mg matrix; no interconnection of fission gas bubbles from fuel particle to fuel particle was observed; the interconnected fission gas bubbles that were observed in the irradiated U-7Mo particles resulted in some transport of solid fission products to the U-7Mo/Mg interface; the presence of microstructural pathways in some U-9.1 Mo particles that could allow for transport of fission gases did not result in the apparent presence of large porosity at the U-7Mo/Mg interface; and, the Mg-Al interaction layers that were present at the Mg matrix/Al 6061 cladding interface exhibited good radiation stability, i.e. no large pores.

  16. A Gamma Radiolysis Study of UO{sub 2}F{sub 2} 0.4H{sub 2}O Using Spent Nuclear Fuel Elements from the High Flux Isotope Reactor

    SciTech Connect

    Icenhour, A.S.

    2002-01-24

    The development of a standard for the safe, long-term storage of {sup 233}U-containing materials resulted in the identification of several needed experimental studies. These studies were largely related to the potential for the generation of unacceptable pressures or the formation of deleterious products during storage of uranium oxides. The primary concern was that these conditions could occur as a result of the radiolysis of residual impurities--specifically fluorides and water-by the high radiation fields associated with {sup 233}U/{sup 232}U-containing materials. This report documents the results from a gamma radiolysis experiment in which UO{sub 2}F{sub 2} {center_dot} 0.4H{sub 2}O was loaded in helium. This experiment was performed using spent nuclear fuel elements from the High Flux Isotope Reactor as the gamma source and was a follow-on to experiments conducted previously. It was found that upon gamma irradiation, the UO{sub 2}F{sub 2} {center_dot} 0.4H{sub 2}O released 0{sub 2} with an initial G(O{sub 2}) = 0.01 molecule O{sub 2}/100 eV and that some of the uranium was reduced from U(VI) to U(IV). The high total dose achieved in the SNF elements was sufficient to reach a damage limit for the UO{sub 2}F{sub 2} {center_dot} 0.4H{sub 2}O. This damage limit, measured in terms of the amount of the U(IV) produced, was found to be about 9 wt%.

  17. Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis

    SciTech Connect

    Vondy, D.R.

    1986-04-01

    Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated.

  18. HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING

    SciTech Connect

    Austin, W.; Brinkley, D.

    2011-10-13

    The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D&D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment

  19. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2009-09-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating eight 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). 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 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 ten years to demonstrate and qualify new 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 the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  20. Semiconductor Chemical Reactor Engineering and Photovoltaic Unit Operations.

    ERIC Educational Resources Information Center

    Russell, T. W. F.

    1985-01-01

    Discusses the nature of semiconductor chemical reactor engineering, illustrating the application of this engineering with research in physical vapor deposition of cadmium sulfide at both the laboratory and unit operations scale and chemical vapor deposition of amorphous silicon at the laboratory scale. (JN)

  1. High flux isotope reactor technical specifications

    SciTech Connect

    Not Available

    1982-04-01

    Technical specifications are presented concerning safety limits and limiting safety system settings; limiting conditions for operation; surveillance requirements; design features; administrative controls; and accidents and anticipated transients.

  2. High Flux Commercial Illumination Solution with Intelligent Controls

    SciTech Connect

    Camil Ghiu

    2012-04-30

    This report summarizes the work performed at OSRAM SYLVANIA under US Department of Energy contract DE-EE0003241 for developing a high efficiency LED-based luminaire. A novel light engine module (two versions: standard and super), power supply and luminaire mechanical parts were designed and tested. At steady-state, the luminaire luminous flux is 3156 lumens (lm), luminous efficacy 97.4 LPW and CRI (Ra) 88 at a correlated color temperature (CCT) of 3507K. When the luminaire is fitted with the super version of the light engine the efficacy reaches 130 LPW. In addition, the luminaire is provided with an intelligent control network capable of additional energy savings. The technology developed during the course of this project has been incorporated into a family of products. Recently, the first product in the family has been launched.

  3. Justification of the new approach to the testing of the candidate ITER materials in fission reactor

    NASA Astrophysics Data System (ADS)

    Nikolaenko, V. A.; Karpukhin, V. I.; Krasikov, E. A.; Kuznetsov, V. N.

    Routine approaches to the candidate ITER materials reactor testing have to be modified by taking into account the large, difference between fusion and fission reactor gamma-quanta fluxes. Recent analyses have shown clear evidence of the relationship between the steel ductile to brittle transition temperature (DBTT) shift and the gamma-quanta flux during the reactor irradiation. For example, if during the reactor irradiation of the 25Cr3NiMo type steel gamma-quanta the flux increases from 5 × 10 11 to 5 × 10 12 sm -2 s -1 to obtain the DBTT=160°C, it is necessary to increase the neutron fluence by one order of magnitude. Simultaneous neutron and gamma irradiation is characterized by an Irradiation Composition Factor (ICF) - neutron flux/gamma-quanta flux. For example, for a fast neutron fluence ( E > 0.5 MeV) of 2.4 × 10 19 cm -2 and 1.6 × 10 19 cm -2 and corresponding values of ICF of 0.4 and 2.1, the DBTT shift is greater for the smaller neutron fluence. Hence, the smaller is the ICF the greater is the gamma-quanta of the nescient defects restoring (annealing) action. For the ITER first wall the ICF is 4, whereas for a nuclear reactor it is only 0.1-0.3. Therefore the new approach to the experimental procedure of the experimental procedure of the canditate ITER materials testing in a fission reactor is justified.

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

    SciTech Connect

    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

  5. Extraterrestrial high energy neutrino fluxes

    NASA Technical Reports Server (NTRS)

    Stecker, F. W.

    1979-01-01

    Using the most recent cosmic ray spectra up to 2x10 to the 20th power eV, production spectra of high energy neutrinos from cosmic ray interactions with interstellar gas and extragalactic interactions of ultrahigh energy cosmic rays with 3K universal background photons are presented and discussed. Estimates of the fluxes from cosmic diffuse sources and the nearby quasar 3C273 are made using the generic relationship between secondary neutrinos and gammas and using recent gamma ray satellite data. These gamma ray data provide important upper limits on cosmological neutrinos. Quantitative estimates of the observability of high energy neutrinos from the inner galaxy and 3C273 above atmospheric background for a DUMAND type detector are discussed in the context of the Weinberg-Salam model with sq sin theta omega = 0.2 and including the atmospheric background from the decay of charmed mesons. Constraints on cosmological high energy neutrino production models are also discussed. It appears that important high energy neutrino astronomy may be possible with DUMAND, but very long observing times are required.

  6. Background radiation measurements at high power research reactors

    SciTech Connect

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

    2015-10-23

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

  7. Background radiation measurements at high power research reactors

    DOE PAGES

    Ashenfelter, J.; Yeh, M.; Balantekin, B.; ...

    2015-10-23

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

  8. Background radiation measurements at high power research reactors

    NASA Astrophysics Data System (ADS)

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

    2016-01-01

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

  9. High-flux solar photon processes

    NASA Astrophysics Data System (ADS)

    Lorents, Donald C.; Narang, Subhash; Huestis, David C.; Mooney, Jack L.; Mill, Theodore; Song, Her-King; Ventura, Susanna

    1992-06-01

    This study was commissioned by the National Renewable Energy Laboratory (NREL) for the purpose of identifying high-flux photoprocesses that would lead to beneficial national and commercial applications. The specific focus on high-flux photoprocesses is based on the recent development by NREL of solar concentrator technology capable of delivering record flux levels. We examined photolytic and photocatalytic chemical processes as well as photothermal processes in the search for processes where concentrated solar flux would offer a unique advantage.

  10. High-flux solar photon processes

    SciTech Connect

    Lorents, D.C.; Narang, S.; Huestis, D.C.; Mooney, J.L.; Mill, T.; Song, H.K.; Ventura, S.

    1992-06-01

    This study was commissioned by the National Renewable Energy Laboratory (NREL) for the purpose of identifying high-flux photoprocesses that would lead to beneficial national and commercial applications. The specific focus on high-flux photoprocesses is based on the recent development by NREL of solar concentrator technology capable of delivering record flux levels. We examined photolytic and photocatalytic chemical processes as well as photothermal processes in the search for processes where concentrated solar flux would offer a unique advantage. 37 refs.

  11. Very high temperature measurements: Applications to nuclear reactor safety tests; Mesures des tres hautes temperatures: Applications a des essais de surete des reacteurs nucleaires

    SciTech Connect

    Parga, Clemente-Jose

    2013-09-27

    This PhD dissertation focuses on the improvement of very high temperature thermometry (1100 deg. C to 2480 deg. C), with special emphasis on the application to the field of nuclear reactor safety and severe accident research. Two main projects were undertaken to achieve this objective: - The development, testing and transposition of high-temperature fixed point (HTFP) metal-carbon eutectic cells, from metrology laboratory precision (±0.001 deg. C) to applied research with a reasonable degradation of uncertainties (±3-5 deg. C). - The corrosion study and metallurgical characterization of Type-C thermocouple (service temp. 2300 deg. C) prospective sheath material was undertaken to extend the survivability of TCs used for molten metallic/oxide corium thermometry (below 2000 deg. C)

  12. Fast Flux Test Facility (FFTF) maintenance provisions

    SciTech Connect

    Marshall, J.L.

    1981-05-01

    The Fast Flux Test Facility (FFTF) was designed with maintainability as a primary parameter, and facilities and provisions were designed into the plant to accommodate the maintenance function. This paper describes the FFTF and its systems. Special maintenance equipment and facilities for performing maintenance on radioactive components are discussed. Maintenance provisions designed into the plant to enhance maintainability are also described.

  13. Fast flux test facility hazards assessment

    SciTech Connect

    Sutton, L.N.

    1994-10-24

    This document establishes the technical basis in support of Emergency Planning Activities for the Fast Flux Test Facility on the Hanford Site. The document represents an acceptable interpretation of the implementing guidance document for DOE Order 5500.3A. Through this document, the technical basis for the development of facility specific Emergency Action Levels and the Emergency Planning Zone is demonstrated.

  14. Energy efficient engine high pressure turbine test hardware detailed design report

    NASA Technical Reports Server (NTRS)

    Halila, E. E.; Lenahan, D. T.; Thomas, T. T.

    1982-01-01

    The high pressure turbine configuration for the Energy Efficient Engine is built around a two-stage design system. Moderate aerodynamic loading for both stages is used to achieve the high level of turbine efficiency. Flowpath components are designed for 18,000 hours of life, while the static and rotating structures are designed for 36,000 hours of engine operation. Both stages of turbine blades and vanes are air-cooled incorporating advanced state of the art in cooling technology. Direct solidification (DS) alloys are used for blades and one stage of vanes, and an oxide dispersion system (ODS) alloy is used for the Stage 1 nozzle airfoils. Ceramic shrouds are used as the material composition for the Stage 1 shroud. An active clearance control (ACC) system is used to control the blade tip to shroud clearances for both stages. Fan air is used to impinge on the shroud casing support rings, thereby controlling the growth rate of the shroud. This procedure allows close clearance control while minimizing blade tip to shroud rubs.

  15. Status report on high fidelity reactor simulation.

    SciTech Connect

    Palmiotti, G.; Smith, M.; Rabiti, C.; Lewis, E.; Yang, W.; Leclere,M.; Siegel, A.; Fischer, P.; Kaushik, D.; Ragusa, J.; Lottes, J.; Smith, B.

    2006-12-11

    This report presents the effort under way at Argonne National Laboratory toward a comprehensive, integrated computational tool intended mainly for the high-fidelity simulation of sodium-cooled fast reactors. The main activities carried out involved neutronics, thermal hydraulics, coupling strategies, software architecture, and high-performance computing. A new neutronics code, UNIC, is being developed. The first phase involves the application of a spherical harmonics method to a general, unstructured three-dimensional mesh. The method also has been interfaced with a method of characteristics. The spherical harmonics equations were implemented in a stand-alone code that was then used to solve several benchmark problems. For thermal hydraulics, a computational fluid dynamics code called Nek5000, developed in the Mathematics and Computer Science Division for coupled hydrodynamics and heat transfer, has been applied to a single-pin, periodic cell in the wire-wrap geometry typical of advanced burner reactors. Numerical strategies for multiphysics coupling have been considered and higher-accuracy efficient methods proposed to finely simulate coupled neutronic/thermal-hydraulic reactor transients. Initial steps have been taken in order to couple UNIC and Nek5000, and simplified problems have been defined and solved for testing. Furthermore, we have begun developing a lightweight computational framework, based in part on carefully selected open source tools, to nonobtrusively and efficiently integrate the individual physics modules into a unified simulation tool.

  16. High-Heat-Flux Cyclic Durability of Thermal and Environmental Barrier Coatings

    NASA Technical Reports Server (NTRS)

    Zhu, Dongming; Ghosn, Louis L.; Miller, Robert A.

    2007-01-01

    Advanced ceramic thermal and environmental barrier coatings will play an increasingly important role in future gas turbine engines because of their ability to protect the engine components and further raise engine temperatures. For the supersonic vehicles currently envisioned in the NASA fundamental aeronautics program, advanced gas turbine engines will be used to provide high power density thrust during the extended supersonic flight of the aircraft, while meeting stringent low emission requirements. Advanced ceramic coating systems are critical to the performance, life and durability of the hot-section components of the engine systems. In this work, the laser and burner rig based high-heat-flux testing approaches were developed to investigate the coating cyclic response and failure mechanisms under simulated supersonic long-duration cruise mission. The accelerated coating cracking and delamination mechanism under the engine high-heat-flux, and extended supersonic cruise time conditions will be addressed. A coating life prediction framework may be realized by examining the crack initiation and propagation in conjunction with environmental degradation under high-heat-flux test conditions.

  17. Advanced propulsion engine assessment based on a cermet reactor

    NASA Technical Reports Server (NTRS)

    Parsley, Randy C.

    1993-01-01

    A preferred Pratt & Whitney conceptual Nuclear Thermal Rocket Engine (NTRE) has been designed based on the fundamental NASA priorities of safety, reliability, cost, and performance. The basic philosophy underlying the design of the XNR2000 is the utilization of the most reliable form of ultrahigh temperature nuclear fuel and development of a core configuration which is optimized for uniform power distribution, operational flexibility, power maneuverability, weight, and robustness. The P&W NTRE system employs a fast spectrum, cermet fueled reactor configured in an expander cycle to ensure maximum operational safety. The cermet fuel form provides retention of fuel and fission products as well as high strength. A high level of confidence is provided by benchmark analysis and independent evaluations.

  18. Critical Heat Flux Phenomena at HighPressure & Low Mass Fluxes: NEUP Final Report Part I: Experiments

    SciTech Connect

    Corradini, Michael; Wu, Qiao

    2015-04-30

    This report is a preliminary document presenting an overview of the Critical Heat Flux (CHF) phenomenon, the High Pressure Critical Heat Flux facility (HPCHF), preliminary CHF data acquired, and the future direction of the research. The HPCHF facility has been designed and built to study CHF at high pressure and low mass flux ranges in a rod bundle prototypical of conceptual Small Modular Reactor (SMR) designs. The rod bundle is comprised of four electrically heated rods in a 2x2 square rod bundle with a prototypic chopped-cosine axial power profile and equipped with thermocouples at various axial and circumferential positions embedded in each rod for CHF detection. Experimental test parameters for CHF detection range from pressures of ~80 – 160 bar, mass fluxes of ~400 – 1500 kg/m2s, and inlet water subcooling from ~30 – 70°C. The preliminary data base established will be further extended in the future along with comparisons to existing CHF correlations, models, etc. whose application ranges may be applicable to the conditions of SMRs.

  19. Test bed ion engine development

    NASA Technical Reports Server (NTRS)

    Aston, G.; Deininger, W. D.

    1984-01-01

    A test bed ion (TBI) engine was developed to serve as a tool in exploring the limits of electrostatic ion thruster performance. A description of three key ion engine components, the decoupled extraction and amplified current (DE-AC) accelerator system, field enhanced refractory metal (FERM) hollow cathode and divergent line cusp (DLC) discharge chamber, whose designs and operating philosophies differ markedly from conventional thruster technology is given. Significant program achievements were: (1) high current density DE-AC accelerator system operation at low electric field stress with indicated feasibility of a 60 mA/sq cm argon ion beam; (2) reliable FERM cathode start up times of 1 to 2 secs. and demonstrated 35 ampere emission levels; (3) DLC discharge chamber plasma potentials negative of anode potential; and (4) identification of an efficient high plasma density engine operating mode. Using the performance projections of this program and reasonable estimates of other parameter values, a 1.0 Newton thrust ion engine is identified as a realizable technology goal. Calculations show that such an engine, comparable in beam area to a J series 30 cm thruster, could, operating on Xe or Hg, have thruster efficiencies as high as 0.76 and 0.78 respectively, with a 100 eV/ion discharge loss.

  20. A New High-Speed Oil-Free Turbine Engine Rotordynamic Simulator Test Rig

    NASA Technical Reports Server (NTRS)

    Howard, Samuel A.

    2007-01-01

    A new test rig has been developed for simulating high-speed turbomachinery rotor systems using Oil-Free foil air bearing technology. Foil air bearings have been used in turbomachinery, primarily air cycle machines, for the past four decades to eliminate the need for oil lubrication. The goal of applying this bearing technology to other classes of turbomachinery has prompted the fabrication of this test rig. The facility gives bearing designers the capability to test potential bearing designs with shafts that simulate the rotating components of a target machine without the high cost of building "make-and-break" hardware. The data collected from this rig can be used to make design changes to the shaft and bearings in subsequent design iterations. This paper describes the new test rig and demonstrates its capabilities through the initial run with a simulated shaft system.

  1. Homogeneous fast-flux isotope-production reactor

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a liquid metal fast breeder reactor. Lithium target material is dissolved in the liquid metal coolant in order to facilitate the production and removal of tritium.

  2. Benchmark Development in Support of Generation-IV Reactor Validation (IRPhEP 2010 Handbook)

    SciTech Connect

    John D. Bess; J. Blair Briggs

    2010-06-01

    The March 2010 edition of the International Reactor Physics Experiment Evaluation Project (IRPhEP) Handbook includes additional benchmark data that can be implemented in the validation of data and methods for Generation IV (GEN-IV) reactor designs. Evaluations supporting sodium-cooled fast reactor (SFR) efforts include the initial isothermal tests of the Fast Flux Test Facility (FFTF) at the Hanford Site, the Zero Power Physics Reactor (ZPPR) 10B and 10C experiments at the Idaho National Laboratory (INL), and the burn-up reactivity coefficient of Japan’s JOYO reactor. An assessment of Russia’s BFS-61 assemblies at the Institute of Physics and Power Engineering (IPPE) provides additional information for lead-cooled fast reactor (LFR) systems. Benchmarks in support of the very high temperature reactor (VHTR) project include evaluations of the HTR-PROTEUS experiments performed at the Paul Scherrer Institut (PSI) in Switzerland and the start-up core physics tests of Japan’s High Temperature Engineering Test Reactor. The critical configuration of the Power Burst Facility (PBF) at the INL which used ternary ceramic fuel, U(18)O2-CaO-ZrO2, is of interest for fuel cycle research and development (FCR&D) and has some similarities to “inert-matrix” fuels that are of interest in GEN-IV advanced reactor design. Two additional evaluations were revised to include additional evaluated experimental data, in support of light water reactor (LWR) and heavy water reactor (HWR) research; these include reactor physics experiments at Brazil’s IPEN/MB-01 Research Reactor Facility and the French High Flux Reactor (RHF), respectively. The IRPhEP Handbook now includes data from 45 experimental series (representing 24 reactor facilities) and represents contributions from 15 countries. These experimental measurements represent large investments of infrastructure, experience, and cost that have been evaluated and preserved as benchmarks for the validation of methods and collection of

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

    SciTech Connect

    Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W.

    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.

  4. Advanced burner test reactor preconceptual design report.

    SciTech Connect

    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

  5. Role of plasma enhanced atomic layer deposition reactor wall conditions on radical and ion substrate fluxes

    SciTech Connect

    Sowa, Mark J.

    2014-01-15

    Chamber wall conditions, such as wall temperature and film deposits, have long been known to influence plasma source performance on thin film processing equipment. Plasma physical characteristics depend on conductive/insulating properties of chamber walls. Radical fluxes depend on plasma characteristics as well as wall recombination rates, which can be wall material and temperature dependent. Variations in substrate delivery of plasma generated species (radicals, ions, etc.) impact the resulting etch or deposition process resulting in process drift. Plasma enhanced atomic layer deposition is known to depend strongly on substrate radical flux, but film properties can be influenced by other plasma generated phenomena, such as ion bombardment. In this paper, the chamber wall conditions on a plasma enhanced atomic layer deposition process are investigated. The downstream oxygen radical and ion fluxes from an inductively coupled plasma source are indirectly monitored in temperature controlled (25–190 °C) stainless steel and quartz reactors over a range of oxygen flow rates. Etch rates of a photoresist coated quartz crystal microbalance are used to study the oxygen radical flux dependence on reactor characteristics. Plasma density estimates from Langmuir probe ion saturation current measurements are used to study the ion flux dependence on reactor characteristics. Reactor temperature was not found to impact radical and ion fluxes substantially. Radical and ion fluxes were higher for quartz walls compared to stainless steel walls over all oxygen flow rates considered. The radical flux to ion flux ratio is likely to be a critical parameter for the deposition of consistent film properties. Reactor wall material, gas flow rate/pressure, and distance from the plasma source all impact the radical to ion flux ratio. These results indicate maintaining chamber wall conditions will be important for delivering consistent results from plasma enhanced atomic layer deposition

  6. Instrumentation to Enhance Advanced Test Reactor Irradiations

    SciTech Connect

    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.

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

    SciTech Connect

    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. Thermal evaluation of uranium silicide miniplates irradiated at high heat flux

    SciTech Connect

    Donna P. Guillen

    2012-09-01

    The Gas Test Loop (GTL)-1 irradiation experiment was conducted in the Advanced Test Reactor (ATR) to assess corrosion performance of proposed booster fuel at heat flux levels ~30% above the design operating condition. Sixteen miniplates fabricated from 25% enriched, high-density (4.8 g U/cm3) U3Si2/Al dispersion fuel with 6061 aluminum cladding were subjected to peak beginning of cycle (BOC) heat fluxes ranging from 411 to 593 W/cm2. No adverse impacts to the miniplates were observed at these high heat flux levels. A detailed finite element model was constructed to calculate temperatures and heat flux for an as-run cycle average effective ATR south lobe power of 25.4 MW(t). Miniplate heat flux levels and fuel, cladding, hydroxide, and coolant–hydroxide interface temperatures were calculated using the average hydroxide thickness on each miniplate measured during post-irradiation examination. The purpose of this study was to obtain a best estimate of the as-run experiment temperatures to aid in establishing acceptable heat flux levels and designing fuel qualification experiments for this fuel type.

  9. In-Situ Creep Testing Capability for the Advanced Test Reactor

    SciTech Connect

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

    2012-09-01

    An instrumented creep testing capability is being developed for specimens irradiated in Pressurized Water Reactor (PWR) coolant conditions at the Advanced Test Reactor (ATR). The test rig has been developed such that samples will be subjected to stresses ranging from 92 to 350 MPa at temperatures between 290 and 370 °C up to at least 2 dpa (displacement per atom). The status of Idaho National Laboratory (INL) efforts to develop the test rig in-situ creep testing capability for the ATR is described. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper reports efforts by INL to evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory (HTTL). Initial data from autoclave tests with 304 stainless steel (304 SS) specimens are reported.

  10. AJ26 Rocket Engine Test

    NASA Video Gallery

    Engineers at NASA’s John C. Stennis Space Center conducts the second in a series of verification tests on an Aerojet AJ26 engine that will power the first stage of the Orbital Sciences Corporatio...

  11. Preliminary engineering report for design of a subscale ejector/diffuser system for high expansion ratio space engine testing

    NASA Technical Reports Server (NTRS)

    Wojciechowski, C. J.; Kurzius, S. C.; Doktor, M. F.

    1984-01-01

    The design of a subscale jet engine driven ejector/diffuser system is examined. Analytical results and preliminary design drawings and plans are included. Previously developed performance prediction techniques are verified. A safety analysis is performed to determine the mechanism for detonation suppression.

  12. The Fast Flux Test Facility shutdown program plan

    SciTech Connect

    Guttenberg, S.; Jones, D.H.; Midgett, J.C.; Nielsen, D.L.

    1995-01-01

    The Fast Flux Test Facility (FFTF) is a 400 MWt sodium-cooled research reactor owned by the US Department of Energy (DOE) and operated by the Westinghouse Hanford Company (WHC) on the Hanford Site in southeastern Washington State. The decision was made by the DOE in December, 1993, to initiate shutdown of the FFTF. This paper describes the FFTF Transition Project Plan (1) (formerly the FFTF Shutdown Program Plan) which provides the strategy, major elements, and project baseline for transitioning the FFTF to an industrially and radiologically safe shutdown condition. The Plan, and its resource loaded schedule, indicate this transition can be achieved in a period of six to seven years at a cost of approximately $359 million. The transition activities include reactor defueling, fuel offload to dry cask storage, sodium drain and reaction, management of sodium residuals, shutdown of auxiliary systems, and preparation of appropriate environmental and regulatory documentation. Completion of these activities will involve resolution of many challenging and unique issues associated with shutdown of a large sodium reactor facility. At the conclusion of these activities, the FFTF will be in a safe condition for turnover to the Hanford Site Environmental Restoration Contractor for a long term surveillance and maintenance phase and decommissioning.

  13. Reference site selection report for the advanced liquid metal reactor at the Idaho National Engineering Laboratory

    SciTech Connect

    Sivill, R.L.

    1990-03-01

    This Reference Site Selection Report was prepared by EG G, Idaho Inc., for General Electric (GE) to provide information for use by the Department of Energy (DOE) in selecting a Safety Test Site for an Advanced Liquid Metal Reactor. Similar Evaluation studies are planned to be conducted at other potential DOE sites. The Power Reactor Innovative Small Module (PRISM) Concept was developed for ALMR by GE. A ALMR Safety Test is planned to be performed on a DOE site to demonstrate features and meet Nuclear Regulatory Commission Requirements. This study considered possible locations at the Idaho National Engineering Laboratory that met the ALMR Prototype Site Selection Methodology and Criteria. Four sites were identified, after further evaluation one site was eliminated. Each of the remaining three sites satisfied the criteria and was graded. The results were relatively close. Thus concluding that the Idaho National Engineering Laboratory is a suitable location for an Advanced Liquid Metal Reactor Safety Test. 23 refs., 13 figs., 9 tabs.

  14. ENGINEERING BULLETIN: BIOLOGICAL TOXICITY TESTING

    EPA Science Inventory

    This Engineering Bulletin is intended to provide site managers with information on ecological assessment and biological toxicity testing, applicability of biological toxicity testing, planning effective biological toxicity assessments, descriptions of test methods, limitations, c...

  15. Reactor target from metal chromium for "pure" high-intensive artificial neutrino source

    NASA Astrophysics Data System (ADS)

    Gavrin, V. N.; Kozlova, Yu. P.; Veretenkin, E. P.; Logachev, A. V.; Logacheva, A. I.; Lednev, I. S.; Okunkova, A. A.

    2017-01-01

    The paper presents the first results of development of manufacturing technology of metallic chromium targets from highly enriched isotope 50Cr for irradiation in a high flux nuclear reactor to obtain a compact high intensity neutrino source with low content of radionuclide impurities and minimum losses of enriched isotope. The main technological stages are the hydrolysis of chromyl fluoride, the electrochemical reduction of metallic chromium, the hot isostatic pressing of chromium powder and the electrical discharge machining of chromium bars. The technological stages of hot isostatic pressing of chromium powder and of electrical discharge machining of Cr rods have been tested.

  16. Diamond Wire Cutting of the Tokamak Fusion Test Reactor

    SciTech Connect

    Keith Rule; Erik Perry; Robert Parsells

    2003-01-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 MeV neutrons. The total tritium content within the vessel is in excess of 7,000 Curies, while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the technology was improved and redesigned for the actual cutting of the vacuum vessel. Ten complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of D&D (Decontamination and Decommissioning) activity.

  17. DIAMOND WIRE CUTTING OF THE TOKAMAK FUSION TEST REACTOR

    SciTech Connect

    Rule, Keith; Perry, Erik; Parsells, Robert

    2003-02-27

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the Tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the techno logy was improved and redesigned for the actual cutting of the vacuum vessel. 10 complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of D&D activity.

  18. Report on the joint meeting of the Division of Development and Technology Plasma/Wall Interaction and High Heat Flux Materials and Components Task Groups

    SciTech Connect

    Wilson, K.L.

    1985-10-01

    This report of the Joint Meeting of the Division of Development and Technology Plasma/Wall Interaction and High Heat Flux Materials and Components Task Groups contains contributing papers in the following areas: Plasma/Materials Interaction Program and Technical Assessment, High Heat Flux Materials and Components Program and Technical Assessment, Pumped Limiters, Ignition Devices, Program Planning Activities, Compact High Power Density Reactor Requirements, Steady State Tokamaks, and Tritium Plasma Experiments. All these areas involve the consideration of High Heat Flux on Materials and the Interaction of the Plasma with the First Wall. Many of the Test Facilities are described as well. (LSP)

  19. Flux flow and flux dynamics in high-Tc superconductors

    NASA Technical Reports Server (NTRS)

    Bennett, L. H.; Turchinskaya, M.; Swartzendruber, L. J.; Roitburd, A.; Lundy, D.; Ritter, J.; Kaiser, D. L.

    1991-01-01

    Because high temperature superconductors, including BYCO and BSSCO, are type 2 superconductors with relatively low H(sub c 1) values and high H(sub c 2) values, they will be in a critical state for many of their applications. In the critical state, with the applied field between H(sub c 1) and H(sub c 2), flux lines have penetrated the material and can form a flux lattice and can be pinned by structural defects, chemical inhomogeneities, and impurities. A detailed knowledge of how flux penetrates the material and its behavior under the influence of applied fields and current flow, and the effect of material processing on these properties, is required in order to apply, and to improve the properties of these superconductors. When the applied field is changed rapidly, the time dependence of flux change can be divided into three regions, an initial region which occurs very rapidly, a second region in which the magnetization has a 1n(t) behavior, and a saturation region at very long times. A critical field is defined for depinning, H(sub c,p) as that field at which the hysteresis loop changes from irreversible to reversible. As a function of temperature, it is found that H(sub c,p) is well described by a power law with an exponent between 1.5 and 2.5. The behavior of H(sub c,p) for various materials and its relationship to flux flow and flux dynamics are discussed.

  20. INEL advanced test reactor plutonium-238 production feasibility assessment

    SciTech Connect

    Schnitzler, B.G. )

    1993-01-10

    Results of a preliminary neutronics assessment indicate the feasibility of [sup 238]Pu production in the Idaho National Engineering Laboratory Advanced Test Reactor (ATR). Based on the results of this assessment, an annual production of 11.3 kg [sup 238]Pu can be achieved in the ATR. An annual loading of 102 kg [sup 237]Np is required for the particular target configuration and irradiation scenario examined. The [sup 236]Pu contaminant level is approximately 6 parts per million at zero cooling time. The product quality is about 90% [sup 238]Pu. Neptunium feedstock requirements, [sup 238]Pu production rates, or product purity can be optimized depending on their relative importances.

  1. INEL advanced test reactor plutonium-238 production feasibility assessment

    NASA Astrophysics Data System (ADS)

    Schnitzler, Bruce G.

    1993-01-01

    Results of a preliminary neutronics assessment indicate the feasibility of 238Pu production in the Idaho National Engineering Laboratory Advanced Test Reactor (ATR). Based on the results of this assessment, an annual production of 11.3 kg 238Pu can be achieved in the ATR. An annual loading of 102 kg 237Np is required for the particular target configuration and irradiation scenario examined. The 236Pu contaminant level is approximately 6 parts per million at zero cooling time. The product quality is about 90% 238Pu. Neptunium feedstock requirements, 238Pu production rates, or product purity can be optimized depending on their relative importances.

  2. EUV Engineering Test Stand

    SciTech Connect

    Tichenor, D.A.; Kubiak, G.D.; Replogle, W.C.; Klebanoff, L.E.; Wronosky, J.B.; Hale, L.C.; Chapman, H.N.; Taylor, J.S.; Folta, J.A.; Montcalm, C.; Hudyma, R.M.

    2000-02-14

    The Engineering Test Stand (ETS) is an EUV laboratory lithography tool. The purpose of the ETS is to demonstrate EUV full-field imaging and provide data required to support production-tool development. The ETS is configured to separate the imaging system and stages from the illumination system. Environmental conditions can be controlled independently in the two modules to maximize EUV throughput and environmental control. A source of 13.4 nm radiation is provided by a laser plasma source in which a YAG laser beam is focused onto a xenon-cluster target. A condenser system, comprised of multilayer-coated mirrors and grazing-incidence mirrors, collects the EUV radiation and directs it onto a-reflecting reticle. A four-mirror, ring-field optical system, having a numerical aperture of 0.1, projects a 4x-reduction image onto the wafer plane. This design corresponds to a resolution of 70nm at a k{sub 1} of 0.52. The ETS is designed to produce full-field images in step: and-scan mode using vacuum-compatible, one-dimension-long-travel magnetically levitated stages for both reticle and wafer. Reticle protection is incorporated into the ETS design. This paper provides a system overview of the ETS design and specifications.

  3. Heavy Water Components Test Reactor Decommissioning - Major Component Removal

    SciTech Connect

    Austin, W.; Brinkley, D.

    2010-05-05

    The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these

  4. Industrial Gas Turbine Engine Catalytic Pilot Combustor-Prototype Testing

    SciTech Connect

    Etemad, Shahrokh; Baird, Benjamin; Alavandi, Sandeep; Pfefferle, William

    2010-04-01

    , Incorporated Saturn engine rig. High pressure single-injector rig and modified engine rig tests demonstrated NOx less than 2 ppm and CO less than 10 ppm over a wide flame temperature operating regime with low combustion noise (<0.15% peak-to-peak). Minimum NOx for the optimized engine retrofit Full RCL® designs was less than 1 ppm with CO emissions less than 10 ppm. Durability testing of the substrate and catalyst material was successfully demonstrated at pressure and temperature showing long term stable performance of the catalytic reactor element. Stable performance of the reactor element was achieved when subjected to durability tests (>5000 hours) at simulated engine conditions (P=15 atm, Tin=400C/750F.). Cyclic tests simulating engine trips was also demonstrated for catalyst reliability. In addition to catalyst tests, substrate oxidation testing was also performed for downselected substrate candidates for over 25,000 hours. At the end of the program, an RCL® catalytic pilot system has been developed and demonstrated to produce NOx emissions of less than 3 ppm (corrected to 15% O2) for 100% and 50% load operation in a production engine operating on natural gas. In addition, a Full RCL® combustor has been designed and demonstrated less than 2 ppm NOx (with potential to achieve 1 ppm) in single injector and modified engine testing. The catalyst/substrate combination has been shown to be stable up to 5500 hrs in simulated engine conditions.

  5. Status of fuel, blanket, and absorber testing in the Fast Flux Test Facility

    SciTech Connect

    Baker, R.B.; Bard, F.E.; Leggett, R.D.; Pitner, A.L.

    1992-11-01

    Over 67,000 fuel, blanket and absorber pins have been irradiated in the Fast Flux Test Facility (FFTF) during its first 12 years of operation. Tests are run in highly controlled and monitored environments with core components similar in size to those in commercial liquid metal reactor (LMR) designs. While primary emphasis was placed on mixed oxide fuels, significant development programs have included metallic fuels, UO[sub 2] blankets, B[sub 4]C absorbers, and other fuels and materials of interest. Irradiation programs for mixed oxides have included progressively lower swelling cladding and duct alloys (e.g., 316 SS, D9 SS, and the ferritic HT9), which also have application to other core components. In many instances the current exposure levels of the advanced FFTF tests are the highest attained and reported in the literature. This paper summarizes the status of irradiation experience at the facility, presents some general conclusions, and reviews the potential for obtaining additional significant data.

  6. Status of fuel, blanket, and absorber testing in the Fast Flux Test Facility

    SciTech Connect

    Baker, R.B.; Bard, F.E.; Leggett, R.D.; Pitner, A.L.

    1992-11-01

    Over 67,000 fuel, blanket and absorber pins have been irradiated in the Fast Flux Test Facility (FFTF) during its first 12 years of operation. Tests are run in highly controlled and monitored environments with core components similar in size to those in commercial liquid metal reactor (LMR) designs. While primary emphasis was placed on mixed oxide fuels, significant development programs have included metallic fuels, UO{sub 2} blankets, B{sub 4}C absorbers, and other fuels and materials of interest. Irradiation programs for mixed oxides have included progressively lower swelling cladding and duct alloys (e.g., 316 SS, D9 SS, and the ferritic HT9), which also have application to other core components. In many instances the current exposure levels of the advanced FFTF tests are the highest attained and reported in the literature. This paper summarizes the status of irradiation experience at the facility, presents some general conclusions, and reviews the potential for obtaining additional significant data.

  7. Thin film heat flux sensor for Space Shuttle Main Engine turbine environment

    NASA Astrophysics Data System (ADS)

    Will, Herbert

    1991-05-01

    The Space Shuttle Main Engine (SSME) turbine environment stresses engine components to their design limits and beyond. The extremely high temperatures and rapid temperature cycling can easily cause parts to fail if they are not properly designed. Thin film heat flux sensors can provide heat loading information with almost no disturbance of gas flows or of the blade. These sensors can provide steady state and transient heat flux information. A thin film heat flux sensor is described which makes it easier to measure small temperature differences across very thin insulating layers.

  8. A High Intensity Multi-Purpose D-D Neutron Generator for Nuclear Engineering Laboratories

    SciTech Connect

    Ka-Ngo Leung; Jasmina L. Vujic; Edward C. Morse; Per F. Peterson

    2005-11-29

    This NEER project involves the design, construction and testing of a low-cost high intensity D-D neutron generator for teaching nuclear engineering students in a laboratory environment without radioisotopes or a nuclear reactor. The neutron generator was designed, fabricated and tested at Lawrence Berkeley National Laboratory (LBNL).

  9. OVERALL CONTROL SYSTEM FOR HIGH FLUX PILE

    DOEpatents

    Newson, H.W.; Durham, N.C.; Wigner, E.P.; Princeton, N.J.; Epler, E.P.

    1961-05-23

    A control system is given for a high fiux reactor incorporating an anti- scram control feature whereby a neutron absorbing control rod acts as a fine adjustment while a neutron absorbing shim rod, actuated upon a command received from reactor period and level signals, has substantially greater effect on the neutron level and is moved prior to scram conditions to alter the reactor activity before a scram condition is created. Thus the probability that a scram will have to be initiated is substantially decreased.

  10. Selecting and using materials for a nuclear rocket engine reactor

    NASA Astrophysics Data System (ADS)

    Lanin, Anatolii G.; Fedik, Ivan I.

    2011-03-01

    This paper provides a historical account of how the nuclear rocket engine reactor was created and discusses the problem of selecting materials for a gas environment at a temperature of up to 3100 K and energy release of 30 MW per liter.

  11. High-pressure hydrogen testing of single crystal superalloys for advanced rocket engine turbopump turbine blades

    NASA Technical Reports Server (NTRS)

    Parr, R. A.; Alter, W. S.; Johnston, M. H.; Strizak, J. P.

    1985-01-01

    A screening program to determine the effects of high pressure hydrogen on selected candidate materials for advanced single crystal turbine blade applications is examined. The alloys chosen for the investigation are CM SX-2, CM SX-4C, Rene N-4, and PWA1480. Testing is carried out in hydrogen and helium at 34 MPa and room temperature, with both notched and unnotched single crystal specimens. Results show a significant variation in susceptibility to Hydrogen Environment Embrittlement (HEE) among the four alloys and a marked difference in fracture topography between hydrogen and helium environment specimens.

  12. High-pressure hydrogen testing of single crystal superalloys for advanced rocket engine turbopump turbine blades

    NASA Technical Reports Server (NTRS)

    Alter, W. S.; Parr, R. A.; Johnston, M. H.; Strizak, J. P.

    1984-01-01

    A screening program to determine the effects of high pressure hydrogen on selected candidate materials for advanced single crystal turbine blade applications is examined. The alloys chosen for the investigation are CM SX-2, CM SX-4C, Rene N-4, and PWA1480. Testing is carried out in hydrogen and helium at 34 MPa and room temperature, with both notched and unnotched single crystal specimens. Results show a significant variation in susceptibility to Hydrogen Environment Embrittlement (HEE) among the four alloys and a marked difference in fracture topography between hydrogen and helium environment specimens.

  13. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    SciTech Connect

    Corwin, William R; Burchell, Timothy D; Katoh, Yutai; McGreevy, Timothy E; Nanstad, Randy K; Ren, Weiju; Snead, Lance Lewis; Wilson, Dane F

    2008-08-01

    materials needed to ensure their safe and reliable operation. The focus of this document will be the overall range of DOE's structural materials research activities being conducted to support VHTR development. By far, the largest portion of material's R&D supporting VHTR development is that being performed directly as part of the Next-Generation Nuclear Plant (NGNP) Project. Supplementary VHTR materials R&D being performed in the DOE program, including university and international research programs and that being performed under direct contracts with the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, will also be described. Specific areas of high-priority materials research that will be needed to deploy the NGNP and provide a basis for subsequent VHTRs are described, including the following: (1) Graphite: (a) Extensive unirradiated materials characterization and assessment of irradiation effects on properties must be performed to qualify new grades of graphite for nuclear service, including thermo-physical and mechanical properties and their changes, statistical variations from billot-to-billot and lot-to-lot, creep, and especially, irradiation creep. (b) Predictive models, as well as codification of the requirements and design methods for graphite core supports, must be developed to provide a basis for licensing. (2) Ceramics: Both fibrous and load-bearing ceramics must be qualified for environmental and radiation service as insulating materials. (3) Ceramic Composites: Carbon-carbon and SiC-SiC composites must be qualified for specialized usage in selected high-temperature components, such as core stabilizers, control rods, and insulating covers and ducting. This will require development of component-specific designs and fabrication processes, materials characterization, assessment of environmental and irradiation effects, and establishment of codes and standards for materials testing and design requirements. (4) Pressure Vessel Steels

  14. Determination of the neutron flux for the Yankee Rowe experiment in the Ford Nuclear Reactor

    SciTech Connect

    Cacciapouti, R.J.; Petrusha, L.

    1994-12-31

    Yankee Atomic Electric Company undertook a Test Irradiation Program at the Ford Nuclear Reactor of the University of Michigan. The program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials. The program was also intended to remove uncertainties in the existing reactor vessel fluence and damage predictions on the Yankee Rowe reactor vessel steel. Since this is the first in-core experiment of this type for the Ford Nuclear Reactor, the measurement of the reaction rate and the estimate of the fluence are presented.

  15. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect

    Not Available

    1988-05-01

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

  16. Testing piezoelectric sensors in a nuclear reactor environment

    NASA Astrophysics Data System (ADS)

    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.

  17. High energy flux thermo-mechanical test of 1D-carbon-carbon fibre composite prototypes for the SPIDER diagnostic calorimeter

    SciTech Connect

    De Muri, M. Pasqualotto, R.; Dalla Palma, M.; Cervaro, V.; Fasolo, D.; Franchin, L.; Tollin, M.; Serianni, G.; Cavallin, T.; Greuner, H.; Böswirth, B.

    2014-02-15

    Operation of the thermonuclear fusion experiment ITER requires additional heating via injection of neutral beams from accelerated negative ions. In the SPIDER test facility, under construction in Padova, the production of negative ions will be studied and optimised. STRIKE (Short-Time Retractable Instrumented Kalorimeter Experiment) is a diagnostic used to characterise the SPIDER beam during short pulse operation (several seconds) to verify if the beam meets the ITER requirements about the maximum allowed beam non-uniformity (below ±10%). The major components of STRIKE are 16 1D-CFC (Carbon-Carbon Fibre Composite) tiles, observed at the rear side by a thermal camera. This contribution gives an overview of some tests under high energy particle flux, aimed at verifying the thermo-mechanical behaviour of several CFC prototype tiles. The tests were performed in the GLADIS facility at IPP (Max-Plank-Institut für Plasmaphysik), Garching. Dedicated linear and nonlinear simulations were carried out to interpret the experiments and a comparison of the experimental data with the simulation results is presented. The results of some morphological and structural studies on the material after exposure to the GLADIS beam are also given.

  18. High energy flux thermo-mechanical test of 1D-carbon-carbon fibre composite prototypes for the SPIDER diagnostic calorimeter.

    PubMed

    De Muri, M; Cavallin, T; Pasqualotto, R; Dalla Palma, M; Cervaro, V; Fasolo, D; Franchin, L; Tollin, M; Greuner, H; Böswirth, B; Serianni, G

    2014-02-01

    Operation of the thermonuclear fusion experiment ITER requires additional heating via injection of neutral beams from accelerated negative ions. In the SPIDER test facility, under construction in Padova, the production of negative ions will be studied and optimised. STRIKE (Short-Time Retractable Instrumented Kalorimeter Experiment) is a diagnostic used to characterise the SPIDER beam during short pulse operation (several seconds) to verify if the beam meets the ITER requirements about the maximum allowed beam non-uniformity (below ±10%). The major components of STRIKE are 16 1D-CFC (Carbon-Carbon Fibre Composite) tiles, observed at the rear side by a thermal camera. This contribution gives an overview of some tests under high energy particle flux, aimed at verifying the thermo-mechanical behaviour of several CFC prototype tiles. The tests were performed in the GLADIS facility at IPP (Max-Plank-Institut für Plasmaphysik), Garching. Dedicated linear and nonlinear simulations were carried out to interpret the experiments and a comparison of the experimental data with the simulation results is presented. The results of some morphological and structural studies on the material after exposure to the GLADIS beam are also given.

  19. High energy flux thermo-mechanical test of 1D-carbon-carbon fibre composite prototypes for the SPIDER diagnostic calorimeter

    NASA Astrophysics Data System (ADS)

    De Muri, M.; Cavallin, T.; Pasqualotto, R.; Dalla Palma, M.; Cervaro, V.; Fasolo, D.; Franchin, L.; Tollin, M.; Greuner, H.; Böswirth, B.; Serianni, G.

    2014-02-01

    Operation of the thermonuclear fusion experiment ITER requires additional heating via injection of neutral beams from accelerated negative ions. In the SPIDER test facility, under construction in Padova, the production of negative ions will be studied and optimised. STRIKE (Short-Time Retractable Instrumented Kalorimeter Experiment) is a diagnostic used to characterise the SPIDER beam during short pulse operation (several seconds) to verify if the beam meets the ITER requirements about the maximum allowed beam non-uniformity (below ±10%). The major components of STRIKE are 16 1D-CFC (Carbon-Carbon Fibre Composite) tiles, observed at the rear side by a thermal camera. This contribution gives an overview of some tests under high energy particle flux, aimed at verifying the thermo-mechanical behaviour of several CFC prototype tiles. The tests were performed in the GLADIS facility at IPP (Max-Plank-Institut für Plasmaphysik), Garching. Dedicated linear and nonlinear simulations were carried out to interpret the experiments and a comparison of the experimental data with the simulation results is presented. The results of some morphological and structural studies on the material after exposure to the GLADIS beam are also given.

  20. Characteristics of Spent Fuel from Plutonium Disposition Reactors, Vol. 1: The Combustion Engineering System 80+ Pressurized-Water-Reactor Design

    SciTech Connect

    Murphy, B.D.

    1993-01-01

    This report discusses a simulation study of the burnup of mixed-oxide fuel in a Combustion Engineering System 80+ Pressurized-Water Reactor. The mixed oxide was composed of uranium and plutonium oxides where the plutonium was of weapons-grade composition. The study was part of the Fissile Materials Disposition Program that considered the possibility of fueling commercial reactors with weapons plutonium. The isotopic composition of the spent fuel is estimated at various times following discharge. Actinides and all significant fission products are considered. The activities, decay-heat values, and gamma-ray fluxes associated with the spent fuel are also discussed. It is clear from the analysis that following discharge the plutonium is no longer of weapons-grade composition. The characteristics of the mixed-oxide fuel at various times following discharge indicate its behavior under long-term storage. As a counterpoint to the mixed-oxide fuel case, the situation with a similar reactor fueled with uranium oxide alone is analyzed. The comparisons serve to emphasize the significance of the plutonium as part of the fuel. For the mixed-oxide case, the burnup was 42,200 MWd/MTHM; in the pure-uranium case, it was 47,800 MWd/MTHM.

  1. Neutron-flux profile monitor for use in a fission reactor

    DOEpatents

    Kopp, M.K.; Valentine, K.H.

    1981-09-15

    A neutron flux monitor is provided which consists of a plurality of fission counters arranged as spaced-apart point detectors along a delay line. As a fission event occurs in any one of the counters, two delayed current pulses are generated at the output of the delay line. The time separation of the pulses identifies the counter in which the particular fission event occurred. Neutron flux profiles of reactor cores can be more accurately measured as a result.

  2. Systems Engineering, Quality and Testing

    NASA Technical Reports Server (NTRS)

    Shepherd, Christena C.

    2015-01-01

    AS9100 has little to say about how to apply a Quality Management System (QMS) to aerospace test programs. There is little in the quality engineering Body of Knowledge that applies to testing, unless it is nondestructive examination or some type of lab or bench testing. If one examines how the systems engineering processes are implemented throughout a test program; and how these processes can be mapped to AS9100, a number of areas for involvement of the quality professional are revealed.

  3. Dielectric Heaters for Testing Spacecraft Nuclear Reactors

    NASA Technical Reports Server (NTRS)

    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.

  4. Reactor technology assessment and selection utilizing systems engineering approach

    SciTech Connect

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In

    2014-02-12

    The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

  5. Reactor technology assessment and selection utilizing systems engineering approach

    NASA Astrophysics Data System (ADS)

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In

    2014-02-01

    The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

  6. Aircraft and Engine Development Testing

    DTIC Science & Technology

    1986-09-01

    Control in Flight * Integrated Inlet- engine * Power/weight Exceeds Unity F-lll * Advanced Engines * Augmented Turbofan * High Turbine Temperature...residence times). Also, fabrication of a small scale "hot" engine with rotating components such as compressors and turbines with cooled blades , is...capabil- ities are essential to meet the needs of current and projected aircraft and engine programs. The required free jet nozzles should be capable of

  7. Concept of an inherently-safe high temperature gas-cooled reactor

    NASA Astrophysics Data System (ADS)

    Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

    2012-06-01

    As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of the confinement function are identified. It is proposed not to apply the probabilistic approach for the evaluation of the radiological consequences of the accidents in the safety analysis because no inherent safety features fail for the mitigation of the consequences of the accidents. Consequently, there are no event sequences to harmful release of radioactive materials if the design extension conditions occur in the inherently-safe HTGR concept. The concept and future R&D items for the inherently-safe HTGR are described in this paper.

  8. Concept of an inherently-safe high temperature gas-cooled reactor

    SciTech Connect

    Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

    2012-06-06

    As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of the confinement function are identified. It is proposed not to apply the probabilistic approach for the evaluation of the radiological consequences of the accidents in the safety analysis because no inherent safety features fail for the mitigation of the consequences of the accidents. Consequently, there are no event sequences to harmful release of radioactive materials if the design extension conditions occur in the inherently-safe HTGR concept. The concept and future R and D items for the inherently-safe HTGR are described in this paper.

  9. Medium-size high-temperature gas-cooled reactor

    SciTech Connect

    Peinado, C.O.; Koutz, S.L.

    1980-08-01

    This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760/sup 0/C (1400/sup 0/F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics (a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant) and engineered safety features (core auxiliary cooling, relief valve, and steam generator dump systems).

  10. Improved measurement of the reactor antineutrino flux and spectrum at Daya Bay

    SciTech Connect

    An, F. P.; Balantekin, A. B.; Band, H. R.; Bishai, M.; Blyth, S.; Cao, D.; Cao, G. F.; Cao, J.; Cen, W. R.; Chan, Y. L.; Chang, J. F.; Chang, L. C.; Chang, Y.; Chen, H. S.; Chen, Q. Y.; Chen, S. M.; Chen, Y. X.; Chen, Y.; Cheng, J. -H.; Cheng, J.; Cheng, Y. P.; Cheng, Z. K.; Cherwinka, J. J.; Chu, M. C.; Chukanov, A.; Cummings, J. P.; de Arcos, J.; Deng, Z. Y.; Ding, X. F.; Ding, Y. Y.; Diwan, M. V.; Dolgareva, M.; Dove, J.; Dwyer, D. A.; Edwards, W. R.; Gill, R.; Gonchar, M.; Gong, G. H.; Gong, H.; Grassi, M.; Gu, W. Q.; Guan, M. Y.; Guo, L.; Guo, R. P.; Guo, X. H.; Guo, Z.; Hackenburg, R. W.; Han, R.; Hans, S.; He, M.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. X.; Huang, X. T.; Huber, P.; Huo, W.; Hussain, G.; Jaffe, D. E.; Jaffke, P.; Jen, K. L.; Jetter, S.; Ji, X. P.; Ji, X. L.; Jiao, J. B.; Johnson, R. A.; Jones, D.; Joshi, J.; Kang, L.; Kettell, S. H.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, J. H. C.; Lei, R. T.; Leitner, R.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, S.; Li, S. C.; Li, W. D.; Li, X. N.; Li, Y. F.; Li, Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, S.; Lin, S. K.; Lin, Y. -C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, D. W.; Liu, J. L.; Liu, J. C.; Loh, C. W.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, K. B.; Lv, Z.; Ma, Q. M.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Malyshkin, Y.; Martinez Caicedo, D. A.; McDonald, K. T.; McKeown, R. D.; Mitchell, I.; Mooney, M.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Ngai, H. Y.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pan, H. -R.; Park, J.; Patton, S.; Pec, V.; Peng, J. C.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Steiner, H.; Sun, G. X.; Sun, J. L.; Tang, W.; Taychenachev, D.; Treskov, K.; Tsang, K. V.; Tull, C. E.; Viaux, N.; Viren, B.; Vorobel, V.; Wang, C. H.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Wei, H. Y.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wise, T.; Wong, H. L. H.; Wong, S. C. F.; Worcester, E.; Wu, C. -H.; Wu, Q.; Wu, W. J.; Xia, D. M.; Xia, J. K.; Xing, Z. Z.; Xu, J. Y.; Xu, J. L.; Xu, Y.; Xue, T.; Yang, C. G.; Yang, H.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Ye, Z.; Yeh, M.; Young, B. L.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, H. H.; Zhang, J. W.; Zhang, Q. M.; Zhang, X. T.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhao, Q. W.; Zhao, Y. B.; Zhong, W. L.; Zhou, L.; Zhou, N.; Zhuang, H. L.; Zou, J. H.

    2017-01-01

    Here, a new measurement of the reactor antineutrino flux and energy spectrum by the Daya Bay reactor neutrino experiment is reported. The antineutrinos were generated by six 2.9 GWth nuclear reactors and detected by eight antineutrino detectors deployed in two near (560 m and 600 m flux-weighted baselines) and one far (1640 m flux-weighted baseline) underground experimental halls. With 621 days of data, more than 1.2 million inverse beta decay (IBD) candidates were detected. The IBD yield in the eight detectors was measured, and the ratio of measured to predicted flux was found to be 0.946 ± 0.020 (0.992 ± 0.021) for the Huber+Mueller (ILL+Vogel) model. A 2.9σ deviation was found in the measured IBD positron energy spectrum compared to the predictions. In particular, an excess of events in the region of 4$-$6 MeV was found in the measured spectrum, with a local significance of 4.4σ. Finally, a reactor antineutrino spectrum weighted by the IBD cross section is extracted for model-independent predictions.

  11. Improved measurement of the reactor antineutrino flux and spectrum at Daya Bay

    DOE PAGES

    An, F. P.; Balantekin, A. B.; Band, H. R.; ...

    2017-01-01

    Here, a new measurement of the reactor antineutrino flux and energy spectrum by the Daya Bay reactor neutrino experiment is reported. The antineutrinos were generated by six 2.9 GWth nuclear reactors and detected by eight antineutrino detectors deployed in two near (560 m and 600 m flux-weighted baselines) and one far (1640 m flux-weighted baseline) underground experimental halls. With 621 days of data, more than 1.2 million inverse beta decay (IBD) candidates were detected. The IBD yield in the eight detectors was measured, and the ratio of measured to predicted flux was found to be 0.946 ± 0.020 (0.992 ±more » 0.021) for the Huber+Mueller (ILL+Vogel) model. A 2.9σ deviation was found in the measured IBD positron energy spectrum compared to the predictions. In particular, an excess of events in the region of 4$-$6 MeV was found in the measured spectrum, with a local significance of 4.4σ. Finally, a reactor antineutrino spectrum weighted by the IBD cross section is extracted for model-independent predictions.« less

  12. Two-field and drift-flux models with applications to nuclear reactor safety

    SciTech Connect

    Travis, J.R.

    1984-05-01

    The ideas of the two-field (6 equation model) and drift-flux (4 equation model) description of two-phase flows are presented. Several example calculations relating to reactor safety are discussed and comparisons of the numerical results and experimental data are shown to be in good agreement.

  13. Flux analysis and metabolomics for systematic metabolic engineering of microorganisms.

    PubMed

    Toya, Yoshihiro; Shimizu, Hiroshi

    2013-11-01

    Rational engineering of metabolism is important for bio-production using microorganisms. Metabolic design based on in silico simulations and experimental validation of the metabolic state in the engineered strain helps in accomplishing systematic metabolic engineering. Flux balance analysis (FBA) is a method for the prediction of metabolic phenotype, and many applications have been developed using FBA to design metabolic networks. Elementary mode analysis (EMA) and ensemble modeling techniques are also useful tools for in silico strain design. The metabolome and flux distribution of the metabolic pathways enable us to evaluate the metabolic state and provide useful clues to improve target productivity. Here, we reviewed several computational applications for metabolic engineering by using genome-scale metabolic models of microorganisms. We also discussed the recent progress made in the field of metabolomics and (13)C-metabolic flux analysis techniques, and reviewed these applications pertaining to bio-production development. Because these in silico or experimental approaches have their respective advantages and disadvantages, the combined usage of these methods is complementary and effective for metabolic engineering.

  14. Advances in process intensification through multifunctional reactor engineering.

    SciTech Connect

    Cooper, Marcia A.; Miller, James Edward; O'Hern, Timothy John; Gill, Walter; Evans, Lindsey R.

    2011-02-01

    A multifunctional reactor is a chemical engineering device that exploits enhanced heat and mass transfer to promote production of a desired chemical, combining more than one unit operation in a single system. The main component of the reactor system under study here is a vertical column containing packing material through which liquid(s) and gas flow cocurrently downward. Under certain conditions, a range of hydrodynamic regimes can be achieved within the column that can either enhance or inhibit a desired chemical reaction. To study such reactors in a controlled laboratory environment, two experimental facilities were constructed at Sandia National Laboratories. One experiment, referred to as the Two-Phase Experiment, operates with two phases (air and water). The second experiment, referred to as the Three-Phase Experiment, operates with three phases (immiscible organic liquid and aqueous liquid, and nitrogen). This report describes the motivation, design, construction, operational hazards, and operation of the both of these experiments. Data and conclusions are included.

  15. Manned engineering test satellites

    NASA Astrophysics Data System (ADS)

    Seko, Hiromi; Satou, Masao; Tomoeda, Hisao; Obara, Hiroaki; Oomura, Katsutoshi

    1992-07-01

    An overview of the conceptual design of manned engineering satellites is presented. The mission scenarios for developing manned engineering satellites involve: (1) selecting mission equipment to enable preferential development and validation of mandatory technologies among those for Environmental Control and Life Support System (ECLSS); (2) selecting mission equipment to enable development and validation of independent domestic technologies as well as utilizing to the utmost the manned space technology acquired through the Japanese Experimental Module (JEM); and (3) installing the most effective mission on the basis of relationships with Extravehicular Activity (EVA), robot, and retrieval-type spacecraft technologies, and trends of overseas manned space technology. The results of reviews on the system and subsystems, such as attitude and orbit control, structure, thermal control, electric power, communication, and data processing subsystems are outlined. The results of reviews on the structure, weight, and reentry phase operation are presented.

  16. Final Assembly and Initial Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect

    S. B. Grover

    2007-05-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) 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 irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new 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.1,2 The experiments, which will each consist of 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 final design phase for the first experiment was completed in 2005, and the fabrication and assembly of the first experiment test train (designated AGR-1) as well as the support systems and fission product monitoring system that will monitor and control the experiment

  17. Tetrakis-amido high flux membranes

    DOEpatents

    McCray, Scott B.

    1989-01-01

    Composite RO membranes of a microporous polymeric support and a polyamide reaction product of a tetrakis-aminomethyl compound and a polyacylhalide are disclosed, said membranes exhibiting high flux and good chlorine resistance.

  18. Tetrakis-amido high flux membranes

    DOEpatents

    McCray, S.B.

    1989-10-24

    Composite RO membranes of a microporous polymeric support and a polyamide reaction product of a tetrakis-aminomethyl compound and a polyacylhalide are disclosed, said membranes exhibiting high flux and good chlorine resistance.

  19. Irradiation performance of Fast Flux Test Facility drivers using D9 alloy

    SciTech Connect

    Pitner, A.L.; Gneiting, B.C.; Bard, F.E.

    1995-11-01

    In comparison with the Fast Flux Test Facility Type 316 stainless steel driver design, six test assemblies employing D9 alloy in place of stainless steel for duct, cladding, and wire wrap material were irradiated to demonstrate the improved performance and lifetime capability of an advanced D9 alloy driver design. A single pinhole-type breach occurred in one of the high-exposure tests after a peak fuel burnup of 155 MWd/kg metal (M) and peak fast neutron fluence of 25 {times} 10{sup 22} n/cm{sup 2} (E > 0.1 MeV). Postirradiation examinations were performed on four of the test assemblies and measured results were compared with analytical evaluations. A revised swelling correlation for D9 alloy was developed to provide improved agreement between calculated and measured cladding deformation results. A fuel pin lifetime design criterion of 5% calculated hoop strain was derived from these results. Alternatively, fuel pin lifetimes were developed for two irradiation parameters using statistical failure analyses. For a 99.99% reliability, the analyses indicated a peak fast-fluence lifetime of 21.0 {times} 10{sup 22} n/cm{sup 2}, or a peak fuel burnup >120 MWd/kg M. In comparison with the Fast Flux Test Facility reference driver design, the extended lifetime capability of D9 alloy would reduce fuel supply requirements for the liquid-metal reactor by a third.

  20. Institute for High Heat Flux Removal (IHHFR). Phases I, II, and III

    SciTech Connect

    Boyd, Ronald D.

    2014-08-31

    The IHHFR focused on interdisciplinary applications as it relates to high heat flux engineering issues and problems which arise due to engineering systems being miniaturized, optimized, or requiring increased high heat flux performance. The work in the IHHFR focused on water as a coolant and includes: (1) the development, design, and construction of the high heat flux flow loop and facility; (2) test section development, design, and fabrication; and, (3) single-side heat flux experiments to produce 2-D boiling curves and 3-D conjugate heat transfer measurements for single-side heated test sections. This work provides data for comparisons with previously developed and new single-side heated correlations and approaches that address the single-side heated effect on heat transfer. In addition, this work includes the addition of single-side heated circular TS and a monoblock test section with a helical wire insert. Finally, the present work includes: (1) data base expansion for the monoblock with a helical wire insert (only for the latter geometry), (2) prediction and verification using finite element, (3) monoblock model and methodology development analyses, and (4) an alternate model development for a hypervapotron and related conjugate heat transfer controlling parameters.

  1. CHARACTERIZATION OF RADIOACTIVITY IN THE REACTOR VESSEL OF THE HEAVY WATER COMPONENT TEST REACTOR

    SciTech Connect

    Vinson, Dennis

    2010-06-01

    The Heavy Water Component Test Reactor (HWCTR) facility is a pressurized heavy water reactor that was used to test candidate fuel designs for heavy water power reactors. The reactor operated at nominal power of 50 MW{sub th}. The reactor coolant loop operated at 1200 psig and 250 C. Two isolated test loop were designed into the reactor to provide special test conditions. Fig. 1 shows a cut-away view of the reactor. The two loops are contained in four inch diameter stainless steel piping. The HWCTR was operated for only a short duration, from March 1962 to December 1964 in order to test the viability of test fuel elements and other reactor components for use in a heavy water power reactor. The reactor achieved 13,882 MWd of total power while testing 36 different fuel assemblies. In the course of operation, HWCTR experienced the cladding failures of 10 separate test fuel assemblies. In each case, the cladding was breached with some release of fuel core material into the isolated test loop, causing fission product and actinide contamination in the main coolant loop and the liquid and boiling test loops. Despite the contribution of the contamination from the failed fuel, the primary source of radioactivity in the HWCTR vessel and internals is the activation products in the thermal shields, and to a lesser degree, activation products in the reactor vessel walls and liner. A detailed facility characterization report of the HWCTR facility was completed in 1996. Many of the inputs and assumptions in the 1996 characterization report were derived from the HWCTR decommissioning plan published in 1975. The current paper provides an updated assessment of the radioisotopic characteristics of the HWCTR vessel and internals to support decommissioning activities on the facility.

  2. Fluoride Salt-Cooled High-Temperature Reactor Development Roadmap

    SciTech Connect

    Holcomb, David Eugene; Flanagan, George F; Mays, Gary T; Pointer, William David; Robb, Kevin R; Yoder Jr, Graydon L

    2014-01-01

    Fluoride salt-cooled high-temperature reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics and fully passive safety. This paper provides an overview of a technology development pathway for expeditious commercial deployment of first-generation FHRs. The paper describes the principal remaining FHR technology challenges and the development path needed to address the challenges. First-generation FHRs do not appear to require any technology breakthroughs, but will require significant technology development and demonstration. FHRs are currently entering early phase engineering development. As such, the development roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant; the lack of an approved licensing framework; the lack of qualified, salt-compatible structural materials; and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

  3. The Need for Cyber-Informed Engineering Expertise for Nuclear Research Reactors

    SciTech Connect

    Anderson, Robert Stephen

    2015-12-01

    Engineering disciplines may not currently understand or fully embrace cyber security aspects as they apply towards analysis, design, operation, and maintenance of nuclear research reactors. Research reactors include a wide range of diverse co-located facilities and designs necessary to meet specific operational research objectives. Because of the nature of research reactors (reduced thermal energy and fission product inventory), hazards and risks may not have received the same scrutiny as normally associated with power reactors. Similarly, security may not have been emphasized either. However, the lack of sound cybersecurity defenses may lead to both safety and security impacts. Risk management methodologies may not contain the foundational assumptions required to address the intelligent adversary’s capabilities in malevolent cyber attacks. Although most research reactors are old and may not have the same digital footprint as newer facilities, any digital instrument and control function must be considered as a potential attack platform that can lead to sabotage or theft of nuclear material, especially for some research reactors that store highly enriched uranium. This paper will provide a discussion about the need for cyber-informed engineering practices that include the entire engineering lifecycle. Cyber-informed engineering as referenced in this paper is the inclusion of cybersecurity aspects into the engineering process. A discussion will consider several attributes of this process evaluating the long-term goal of developing additional cyber safety basis analysis and trust principles. With a culture of free information sharing exchanges, and potentially a lack of security expertise, new risk analysis and design methodologies need to be developed to address this rapidly evolving (cyber) threatscape.

  4. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    SciTech Connect

    Michael Swanson; Daniel Laudal

    2008-03-31

    -reactivity (low-rank) coals appear to perform better in a transport reactor than the less reactive bituminous coals. Factors that affect TRDU product gas quality appear to be coal type, temperature, and air/coal ratios. Testing with a higher-ash, high-moisture, low-rank coal from the Red Hills Mine of the Mississippi Lignite Mining Company has recently been completed. Testing with the lignite coal generated a fuel gas with acceptable heating value and a high carbon conversion, although some drying of the high-moisture lignite was required before coal-feeding problems were resolved. No ash deposition or bed material agglomeration issues were encountered with this fuel. In order to better understand the coal devolatilization and cracking chemistry occurring in the riser of the transport reactor, gas and solid sampling directly from the riser and the filter outlet has been accomplished. This was done using a baseline Powder River Basin subbituminous coal from the Peabody Energy North Antelope Rochelle Mine near Gillette, Wyoming.

  5. PBF Reactor Building (PER620). Detail of fuel test assembly in ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    PBF Reactor Building (PER-620). Detail of fuel test assembly in preparation for test. When complete, it will fit into in-pile tube. The maximum outside diameter of which must be about 8.25 inches. Date: 1982. INEEL negative no. 82-4908 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  6. Engine Certification and Compliance Testing

    EPA Pesticide Factsheets

    The National Vehicle and Fuel Emissions Laboratory (NVFEL) tests a portion of all heavy-duty diesel and small gasoline engines intended for sale in the United States to confirm compliance with EPA’s exhaust emissions standards.

  7. A fission matrix based validation protocol for computed power distributions in the advanced test reactor

    SciTech Connect

    Nielsen, J. W.; Nigg, D. W.; LaPorta, A. W.

    2013-07-01

    The Idaho National Laboratory (INL) has been engaged in a significant multi year effort to modernize the computational reactor physics tools and validation procedures used to support operations of the Advanced Test Reactor (ATR) and its companion critical facility (ATRC). Several new protocols for validation of computed neutron flux distributions and spectra as well as for validation of computed fission power distributions, based on new experiments and well-recognized least-squares statistical analysis techniques, have been under development. In the case of power distributions, estimates of the a priori ATR-specific fuel element-to-element fission power correlation and covariance matrices are required for validation analysis. A practical method for generating these matrices using the element-to-element fission matrix is presented, along with a high-order scheme for estimating the underlying fission matrix itself. The proposed methodology is illustrated using the MCNP5 neutron transport code for the required neutronics calculations. The general approach is readily adaptable for implementation using any multidimensional stochastic or deterministic transport code that offers the required level of spatial, angular, and energy resolution in the computed solution for the neutron flux and fission source. (authors)

  8. NEUTRONIC REACTOR HAVING LOCALIZED AREAS OF HIGH THERMAL NEUTRON DENSITIES

    DOEpatents

    Newson, H.W.

    1958-06-01

    A nuclear reactor for the irradiation of materials designed to provide a localized area of high thermal neutron flux density in which the materials to be irradiated are inserted is described. The active portion of the reactor is comprised of a cubicle graphite moderator of about 25 feet in length along each axis which has a plurality of cylindrical channels for accommodatirg elongated tubular-shaped fuel elements. The fuel elements have radial fins for spacing the fuel elements from the channel walls, thereby providing spaces through which a coolant may be passed, and also to serve as a heatconductirg means. Ducts for accommnodating the sample material to be irradiated extend through the moderator material perpendicular to and between parallel rows of fuel channels. The improvement is in the provision of additional fuel element channels spaced midway between 2 rows of the regular fuel channels in the localized area surrounding the duct where the high thermal neutron flux density is desired. The fuel elements normally disposed in the channels directly adjacent the duct are placed in the additional channels, and the channels directly adjacent the duct are plugged with moderator material. This design provides localized areas of high thermal neutron flux density without the necessity of providing additional fuel material.

  9. Metabolite Valves: Dynamic Control of Metabolic Flux for Pathway Engineering

    NASA Astrophysics Data System (ADS)

    Prather, Kristala

    2015-03-01

    Microbial strains have been successfully engineered to produce a wide variety of chemical compounds, several of which have been commercialized. As new products are targeted for biological synthesis, yield is frequently considered a primary driver towards determining feasibility. Theoretical yields can be calculated, establishing an upper limit on the potential conversion of starting substrates to target compounds. Such yields typically ignore loss of substrate to byproducts, with the assumption that competing reactions can be eliminated, usually by deleting the genes encoding the corresponding enzymes. However, when an enzyme encodes an essential gene, especially one involved in primary metabolism, deletion is not a viable option. Reducing gene expression in a static fashion is possible, but this solution ignores the metabolic demand needed for synthesis of the enzymes required for the desired pathway. We have developed Metabolite valves to address this challenge. The valves are designed to allow high flux through the essential enzyme during an initial period where growth is favored. Following an external perturbation, enzyme activity is then reduced, enabling a higher precursor pool to be diverted towards the pathway of interest. We have designed valves with control at both the transcriptional and post-translational levels. In both cases, key enzymes in glucose metabolism are regulated, and two different compounds are targeted for heterologous production. We have measured increased concentrations of intracellular metabolites once the valve is closed, and have demonstrated that these increased pools lead to increased product yields. These metabolite valves should prove broadly useful for dynamic control of metabolic flux, resulting in improvements in product yields.

  10. Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

    SciTech Connect

    Tong, H; Snow, G C; Chu, E K :; Chang, R L.S.; Angwin, M J; Pessagno, S L

    1981-09-01

    An experimental program was performed to develop durable catalytic reactors for advanced gas turbine engines. This program was performed as part of DOE's Gas Turbine Highway Vehicle Systems Project. Objectives of this program were to evaluate furnace aging as a cost-effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1000 h of combustion durability, and define a catalytic reactor system with a high probability of successfful integration into an automotive gas turbine engine. In the first phase of this program, 14 different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel at 1700 K combustion coditions. The durability reactor, a proprietary UOP noble metal catalyst, failed structurally after about 136 h and the catalyst was essentially inactive after about 226 h. In Phase II, eight additional catalytic reactors were evalated and one of these was sucessfully combustion-tested for 1000 h at 1700 K on propane fuel. This durability reactor used graded-cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

  11. HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

    DOEpatents

    Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

    1960-06-14

    A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

  12. ENGINEERING DEVELOPMENT OF SLURRY BUBBLE COLUMN REACTOR (SBCR) TECHNOLOGY

    SciTech Connect

    Bernard A. Toseland

    2000-12-31

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column reactor to maximize reactor productivity, (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions, and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large-diameter reactors. Washington University's work during the reporting period involved the implementation of the automated calibration device, which will provide an advanced method of determining liquid and slurry velocities at high pressures. This new calibration device is intended to replace the original calibration setup, which depended on fishing lines and hooks to position the radioactive particle. The report submitted by Washington University contains a complete description of the new calibration device and its operation. Improvements to the calibration program are also discussed. Iowa State University utilized air-water bubble column simulations in an effort to determine the domain size needed to represent all of the flow scales in a gas-liquid column at a high superficial velocity. Ohio State's report summarizes conclusions drawn from the completion of gas injection phenomena studies, specifically with respect to the characteristics of bubbling-jetting at submerged single orifices in liquid-solid suspensions.

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

    SciTech Connect

    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

  14. High flux photon beam monitor

    SciTech Connect

    Mortazavi, P.; Woodle, M.; Rarback, H.; Shu, D.; Howells, M.

    1985-01-01

    We have designed two photon beam position monitors for use on our x-ray storage ring beam lines. In both designs, a pair of tungsten blades, separated by a pre-determined gap, intercepts a small fraction of the incoming beam. Due to photoemission, an electrical signal is generated which is proportional to the amount of beam intercepted. The thermal load deposited in the blade is transferred by a heat pipe to a heat exchanger outside the vacuum chamber. A prototype monitor with gap adjustment capability was fabricated and tested at a uv beam line. The results show that the generated electrical signal is a good measurement of the photon beam position. In the following sections, design features and test results are discussed.

  15. New results from pulse tests in the CABRI reactor

    SciTech Connect

    Schmitz, F.; Papin, J.; Haessler, M.

    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.

  16. Engineering tool for trapped proton flux anisotropy evaluation.

    PubMed

    Kruglanski, M

    1996-11-01

    At low altitudes, the high-energy trapped proton fluxes are strongly anisotropic. The proton flux is controlled by the density distribution of the Earth's atmosphere that induces a steep pitch-angle distribution and an East-West effect, the latter caused by the finite size of the proton gyration radius. We have developed a software package, ANISO, to evaluate averaged energy spectra of trapped proton unidirectional fluxes for a given spacecraft orbit and attitude by deducing the angular-dependent proton flux spectra from the AP-8 omnidirectional