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Sample records for high flux engineering test reactor

  1. 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.

  2. 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.

  3. The CG1 instrument development test station at the high flux isotope reactor

    NASA Astrophysics Data System (ADS)

    Crow, Lowell; Robertson, Lee; Bilheux, Hassina; Fleenor, Mike; Iverson, Erik; Tong, Xin; Stoica, Ducu; Lee, W. T.

    2011-04-01

    The CG1 instrument development station at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory began commissioning operation in 2009. When completed, the station will have four beams. CG1A is a 4.22 Å monochromatic beam intended for spin-echo resolved grazing incidence scattering (SERGIS) prototype development. Initial beam operation and characterization are in progress. CG1B will be a 2.35 Å monochromatic beam for a 2-axis utility diffractometer for sample alignment and monochromator development. CG1C will have a double-bounce monochromator system, which will produce a variable wavelength beam from about 1.8-6.4 Å, and will be used for imaging and optical development. The CG1D beam is a single chopper time-of-flight system, used for instrument prototype and component testing. The cold neutron spectrum, with an integrated flux of about 2.7×109 n/cm2 s, has a guide cutoff at 0.8 Å and useful wavelengths greater than 20 Å.Initial results from CG1 include spectral characterization, imaging tests, detector trials, and polarizer tests. An overview of recent tests will be presented, and upcoming instrument prototype efforts will be described.

  4. 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.

  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. Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

    SciTech Connect

    Hauptman, H.M.; Petro, J.N.; Jacobi, O.

    1995-04-01

    This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed.

  7. 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.

  8. 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.

  9. Assessment of similarity of HFBR (High Flux Beam Reactor) with separate effects test

    SciTech Connect

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

    1990-11-01

    A Separate Effects Test (SET) facility was constructed in 1963 to demonstrate the feasibility of the HFBR design and to determine the core power limits for a safe flow reversal event. The objective of the task reported here is to review the capability of the test to scale the dominant phenomena in the HFBR during a flow reversal event and the applicability of the range of the power level obtained from the test to the HFBR. The conclusion of this report was that the flow during the flow reversal event will not be similar in the two facilities. The causes of the dissimilarity are the differences in the core inlet friction, bypass path friction, the absence of the check valve in the test, and the materials used to represent the fuel plates. The impact of these differences is that the HFBR will undergo flow reversal sooner than the test and will have a higher flow rate in the final Natural Circulation Period. The shorter duration of the flow reversal event will allow less time for the plate to heat up and the larger flow in the Natural Circulation Period will lead to higher critical heat flux limits in the HFBR than in the test. Based on these observations, it was concluded that the HFBR can undergo flow reversal safely for heat fluxes up to 46,700 (BTU/hr ft{sup 2}), the heat flux limit obtained from the 1963 test.

  10. 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.

  11. Benchmark analysis of high temperature engineering test reactor core using McCARD code

    SciTech Connect

    Jeong, Chang Joon; Jo, Chang Keun; Lee, Hyun Chul; Noh, Jae Man

    2013-07-01

    A benchmark calculation has been performed for a startup core physics test of Japan's High Temperature Engineering Test Reactor (HTTR). The calculation is carried out by the McCARD code, which adopts the Monte Carlo method. The cross section library is ENDF-B/VII.0. The fuel cell is modeled by the reactivity-equivalent physical transform (RPT) method. Effective multiplication factors with different numbers of fuel columns have been analyzed. The calculation shows that the HTTR becomes critical with 19 fuel columns with an excess reactivity of 0.84% Δk/k. The discrepancies between the measurements and Monte Carlo calculations are 2.2 and 1.4 % Δk/k for 24 and 30 columns, respectively. The reasons for the discrepancy are thought to be the current version of cross section library and the impurity in the graphite which is represented by the boron concentration. In the future, the depletion results will be proposed for further benchmark calculations. (authors)

  12. The High Flux Beam Reactor at Brookhaven National Laboratory

    SciTech Connect

    Shapiro, S.M.

    1994-12-31

    Brookhaven National Laboratory`s High Flux Beam Reactor (HFBR) was built because of the need of the scientist to always want `more`. In the mid-50`s the Brookhaven Graphite reactor was churning away producing a number of new results when the current generation of scientists, led by Donald Hughes, realized the need for a high flux reactor and started down the political, scientific and engineering path that led to the BFBR. The effort was joined by a number of engineers and scientists among them, Chemick, Hastings, Kouts, and Hendrie, who came up with the novel design of the HFBR. The two innovative features that have been incorporated in nearly all other research reactors built since are: (i) an under moderated core arrangement which enables the thermal flux to peak outside the core region where beam tubes can be placed, and (ii) beam tubes that are tangential to the core which decrease the fast neutron background without affecting the thermal beam intensity. Construction began in the fall of 1961 and four years later, at a cost of $12 Million, criticality was achieved on Halloween Night, 1965. Thus began 30 years of scientific accomplishments.

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. 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

  19. High flux research reactors based on particulate fuel

    SciTech Connect

    Powell, J.R.; Takahashi, H.; Horn, F.L.

    1986-02-01

    High Flux Particle Bed Reactor (HFPBR) designs based on High Temperature Gas Reactors (HTGR) particular fuel are described. The coated fuel particles, approx.500 microns in diameter, are packed between porous metal frits, and directly cooled by flowing D/sub 2/O. The large heat transfer surface area in the packed bed, approx.100 cm/sup 2//cm/sup 3/ of volume, allows high power densities, typically 10 MW/liter. Peak thermal fluxes in the HFPBR are 1 to 2 x 1/sup 16/ n/c/sup 2/ sec., depending on configuration and moderator choice with beryllium and D/sub 2/O Moderators yielding the best flux performance. Spent fuel particles can be hydraulically unloaded every day or two and fresh fuel reloaded. The short fuel cycle allows HFPBR fuel loading to be very low, approx.2 kg of /sup 235/U, with a fission product inventory one-tenth of that in present high flux research reactors. The HFPBR can use partially enriched fuel, 20% /sup 235/U, without degradation in flux reactivity. 8 refs., 12 figs., 2 tabs.

  20. Neutron diffraction facilities at the high flux reactor, Petten

    NASA Astrophysics Data System (ADS)

    Ohms, C.; Youtsos, A. G.; Bontenbal, A.; Mulder, F. M.

    2000-03-01

    The High Flux Reactor in Petten is equipped with twelve beam tubes for the extraction of thermal neutrons for applications in materials and medical science. Beam tubes HB4 and HB5 are equipped with diffractometers for residual stress and powder investigations. Recently at HB4 the Large Component Neutron Diffraction Facility has been installed. It is a unique facility with respect to its capability of handling heavy components up to 1000 kg in residual stress testing. Its basic features are described and the first applications on thick piping welds are shown. The diffractometer at HB5 can be set up for powder and stress measurements. Recent applications include temperature dependent measurements on phase transitions in intermetallic compounds and on Li ion energy storage materials.

  1. Neutronic calculations for a new high flux reactor

    SciTech Connect

    Difilippo, F.C.; Worley, B.A.; Vondy, D.R.

    1985-01-01

    The Oak Ridge National Laboratory has begun the design of a new high flux reactor to be used for basic research, isotope production, and material irradiation. One of the principal goals of the design is the production of a thermal flux peak in the reflector larger than 5 x 10/sup 15//cm/sup 2/sec. A theoretical analysis of the slowing-down and diffusion of neutrons produced by a spherical fission source immersed in a moderator shows that the flux per unit power is maximized by combining a very undermoderated core with a very low absorbing reflector. The theoretical model interrelates total power, power density and transport properties with the thermal flux allowing very inexpensive scoping calculations. Full scale and detailed calculations were made with a numerical model which uses the Bold Venture code system. Calculations show that a highly enriched /sup 235/U reactor with D/sub 2/O as moderator and reflector would produce the desired peak flux, and the reactor would have a reasonable core life.

  2. 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.

  3. FFTF (Fast Flux Test Facility) reactor shutdown system reliability reevaluation

    SciTech Connect

    Pierce, B.F.

    1986-07-01

    The reliability analysis of the Fast Flux Test Facility reactor shutdown system was reevaluated. Failure information based on five years of plant operating experience was used to verify original reliability numbers or to establish new ones. Also, system modifications made subsequent to performance of the original analysis were incorporated into the reevaluation. Reliability calculations and sensitivity analyses were performed using a commercially available spreadsheet on a personal computer. The spreadsheet was configured so that future failures could be tracked and compared with expected failures. A number of recommendations resulted from the reevaluation including both increased and decreased surveillance intervals. All recommendations were based on meeting or exceeding existing reliability goals. Considerable cost savings will be incurred upon implementation of the recommendations.

  4. High flux Particle Bed Reactor systems for rapid transmutation of actinides and long lived fission products

    SciTech Connect

    Powell, J.; Ludewig, H.; Maise, G.; Steinberg, M.; Todosow, M.

    1993-08-01

    An initial assessment of several actinide/LLFP burner concepts based on the Particle Bed Reactor (PBR) is described. The high power density/flux level achievable with the PBR make it an attractive candidate for this application. The PBR based actinide burner concept also possesses a number of safety and economic benefits relative to other reactor based transmutation approaches including a low inventory of radionuclides, and high integrity, coated fuel particles which can withstand extremely high in temperatures while retaining virtually all fission products. In addition the reactor also posesses a number of ``engineered safety features,`` which, along with the use of high temperature capable materials further enhance its safety characteristics.

  5. 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)

  6. Performance and Safety Parameters for the High Flux Isotope Reactor

    SciTech Connect

    Ilas, Germina; Primm, Trent

    2012-01-01

    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 ENDV/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared when available 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.

  7. Study of fueling requirements for the Engineering Test Reactor

    SciTech Connect

    Ho, S.K.; Perkins, L.J.

    1987-10-16

    An assessment of the fueling requirement for the TIBER Engineering Test Reactor is studied. The neutral shielding pellet ablation model with the inclusion of the effects of the alpha particles is used for our study. The high electron temperature in a reactor-grade plasma makes pellet penetration very difficult. The launch length has to be very large (several tens of meters) in order to avoid pellet breakage due to the low inertial strength of DT ''ice.'' The minimum repetition rate corresponding to the largest allowable pellet, is found to be about 1 Hz. A brief survey is done on the various operational and conceptual pellet injection schemes for plasma fueling. The underlying conclusion is that an alternative fueling scheme of coaxial compact-toroid plasma gun is very likely needed for effective central fueling of reactor-grade plasmas. 16 refs.

  8. 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.

  9. Initial results of systems analysis of the ETR/ITER (Engineering Test Reactor/International Thermonuclear Experimental Reactor) design space

    SciTech Connect

    Peng, Y.K.M.; Galambos, J.D.; Reid, R.L.; Strickler, D.J.; Kalsi, S.; Deleanu, L.

    1987-01-01

    Preliminary versions of the Engineering Test Reactor (ETR) systems code TETRA (Tokamak Engineering Test Reactor Analysis), which determines design solutions by the method of constrained optimization, are used to characterize the International Thermonuclear Experimental Reactor (ITER) and its design parameter space. We find that the physics objectives of high ignition margin and high plasma current lead to minimum size at relatively low aspect ratios (A = 2.5-3.0), while the engineering objective of high neutron wall load (W/sub L/ /approx gt/ 1.0 MW/m/sup 2/) leads to minimum size at higher A (/approximately/3.5). For minimum-size ITERs, the optimal toroidal field coil (TFC) designs fall within a narrow range of maximum fields (10-11 T) with R varying over only a few percent despite a factor of two change in the winding pack current density J/sub wp/. The major radius of the design is found to be sensitive to changes in elongation, inboard distances (such as plasma scrape-off), inductive flux capability, plasma temperature, beta limit, and ignition margin. A preliminary characterization of the US ITER designs with plasma current I/sub p/ > 15 MA and R < 4.5 m has been obtained by combining the engineering assumptions for devices such as the Tokamak Ignition/Burn Engineering Reactor (TIBER) with the physics assumptions for devices such as the Compact Ignition Tokamak (CIT) and the Next European Torus (NET). These devices can accommodate a range of full- to reduced-bore, driven (Q < 10), steady-state plasmas for the engineering phase that produces high neutron wall load and fluence. 12 refs., 4 figs., 3 tabs.

  10. 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.

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

    SciTech Connect

    HOLDEN,N.E.

    1999-09-10

    RADIATION DOSIMETRY MEASUREMENTS HAVE BEEN PERFORMED OVER A PERIOD OF MANY YEARS AT THE HIGH FLUX BEAM REACTOR (HFBR) AND THE MEDICAL RESEARCH REACTOR (BMRR) AT BROOKHAVEN NATIONAL LABORATORY TO PROVIDE INFORMATION ON THE ENERGY DISTRIBUTION OF THE NEUTRON FLUX, NEUTRON DOSE RATES, GAMMA-RAY FLUXES AND GAMMA-RAY DOSE RATES. THE MCNP PARTICLE TRANSPORT CODE PROVIDED MONTE CARLO RESULTS TO COMPARE WITH VARIOUS DOSIMETRY MEASUREMENTS PERFORMED AT THE EXPERIMENTAL PORTS, AT THE TREATMENT ROOMS AND IN THE THIMBLES AT BOTH HFBR AND BMRR.

  12. 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.

  13. 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.

  14. 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.

  15. 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…

  16. 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.

  17. Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel

    SciTech Connect

    Fauzia, A. F.; Waris, A.; Novitrian

    2010-06-22

    High temperature engineering test reactor (HTTR) is an uranium oxide (UO2) fuel, graphite moderator and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 950 deg. C. Instead of using helium gas, we have utilized carbon dioxide as a coolant in the present study. Beside that, uranium and plutonium oxide (mixed oxide, MOX) and minor actinides have been employed as a new fuel type of HTTR. Utilization of plutonium and minor actinide is one of the support system to non-proliferation issue in the nuclear development. The enrichment for uranium oxide has been varied of 6-20% with plutonium and minor actinides concentration of 10%. In this study, burnup period is 1100 days. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Reactor core calculation was done by using CITATION module. The result shows that HTTR can achieve its criticality condition with 14% of {sup 235}U enrichment.

  18. Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel

    NASA Astrophysics Data System (ADS)

    Fauzia, A. F.; Waris, A.; Novitrian

    2010-06-01

    High temperature engineering test reactor (HTTR) is an uranium oxide (UO2) fuel, graphite moderator and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 950° C. Instead of using helium gas, we have utilized carbon dioxide as a coolant in the present study. Beside that, uranium and plutonium oxide (mixed oxide, MOX) and minor actinides have been employed as a new fuel type of HTTR. Utilization of plutonium and minor actinide is one of the support system to non-proliferation issue in the nuclear development. The enrichment for uranium oxide has been varied of 6-20% with plutonium and minor actinides concentration of 10%. In this study, burnup period is 1100 days. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Reactor core calculation was done by using CITATION module. The result shows that HTTR can achieve its criticality condition with 14% of 235 U enrichment.

  19. Fast Flux Test Facility Reactor Vessel Removal Study

    SciTech Connect

    BOWMAN, B.R.

    2002-10-23

    This study assesses the feasibility of removing the FFTF reactor vessel from its current location in the reactor cavity inside the Containment vessel to a transporter for relocation to a burial pit in the 200 Area.

  20. Testing of actively cooled high heat flux mock-ups

    NASA Astrophysics Data System (ADS)

    Rödig, M.; Duwe, R.; Kühnlein, W.; Linke, J.; Scheerer, M.; Smid, I.; Wiechers, B.

    1998-10-01

    Several un-irradiated CFC monoblock mock-ups have been loaded in thermal fatigue tests up to 1000 cycles at power densities <25 MW/m 2. No indication of failure was observed for these loading conditions. Two of the mock-ups were inspected by ultra-sonic methods before thermal cycling. It could be proved that the voids found in the post-mortem metallography existed before and had no effect on the integrity of the mock-up. For the first time, neutron-irradiated CFC monoblock mock-ups have been tested in the electron beam facility JUDITH. These mock-ups had been irradiated before in the High Flux Reactor at Petten up to 0.3 dpa at 320°C and 770°C. All samples showed a significant increase of surface temperature, due to the irradiation induced decrease in thermal conductivity of the CFC materials.

  1. Experimental study of thermal crisis in connection with Tokamak reactor high heat flux components

    NASA Astrophysics Data System (ADS)

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

    2000-04-01

    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.

  2. 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.

  3. 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.

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

    DOE PAGESBeta

    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

  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. 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

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. Seismic, high wind, tornado, and probabilistic risk assessments of the High Flux Isotope Reactor

    SciTech Connect

    Harris, S.P.; Stover, R.L.; Hashimoto, P.S.; Dizon, J.O.; Oak Ridge National Lab., TN; EQE, Inc., San Francisco, CA )

    1989-01-01

    Natural phenomena analyses were performed on the High Flux Isotope Reactor (HFIR) Deterministic and probabilistic evaluations were made to determine the risks resulting from earthquakes, high winds, and tornadoes. Analytic methods in conjunction with field evaluations and an earthquake experience data base evaluation methods were used to provide more realistic results in a shorter amount of time. Plant modifications completed in preparation for HFIR restart and potential future enhancements are discussed. 5 figs.

  15. 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

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Critical Heat Flux Experiments on the Reactor Vessel Wall Using 2-D Slice Test Section

    SciTech Connect

    Jeong, Yong Hoon; Chang, Soon Heung; Baek, Won-Pil

    2005-11-15

    The critical heat flux (CHF) on the reactor vessel outer wall was measured using the two-dimensional slice test section. The radius and the channel area of the test section were 2.5 m and 10 cm x 15 cm, respectively. The flow channel area and the heater width were smaller than those of the ULPU experiments, but the radius was greater than that of the ULPU. The CHF data under the inlet subcooling of 2 to 25 deg. C and the mass flux 0 to 300 kg/m{sup 2}.s had been acquired. The measured CHF value was generally slightly lower than that of the ULPU. The difference possibly comes from the difference of the test section material and the thickness. However, the general trend of CHF according to the mass flux was similar with that of the ULPU. The experimental CHF data were compared with the predicted values by SULTAN correlation. The SULTAN correlation predicted well this study's data only for the mass flux higher than 200 kg/m{sup 2}.s, and for the exit quality lower than 0.05. The local condition-based correlation was developed, and it showed good prediction capability for broad quality (-0.01 to 0.5) and mass flux (<300 kg/m{sup 2}.s) conditions with a root-mean-square error of 2.4%. There were increases in the CHF with trisodium phosphate-added water.

  1. Three-dimensional calculations of neutron streaming in the beam tubes of the ORNL HFIR (High Flux Isotope Reactor) Reactor

    SciTech Connect

    Childs, R.L.; Rhoades, W.A.; Williams, L.R.

    1988-01-01

    The streaming of neutrons through the beam tubes in High Flux Isotope Reactor at Oak Ridge National Laboratory has resulted in a reduction of the fracture toughness of the reactor vessel. As a result, an evaluation of vessel integrity was undertaken in order to determine if the reactor can be operated again. As a part of this evaluation, three-dimensional neutron transport calculations were performed to obtain fluxes at points of interest in the wall of the vessel. By comparing the calculated and measured activation of dosimetry specimens from the vessel surveillance program, it was determined that the calculated flux shape was satisfactory to transpose the surveillance data to the locations in the vessel. A bias factor was applied to correct for the average C/E ratio of 0.69. 8 refs., 7 figs., 3 tabs.

  2. 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.

  3. High-flux first-wall design for a small reversed-field pinch reactor

    NASA Astrophysics Data System (ADS)

    Cort, G. E.; Graham, A. L.; Christensen, K. E.

    To achieve the goal of a commercially economical fusion power reactor, small physical size and high power density should be combined with simplicity (minimized use of high technology systems). The Reversed-Field Pinch (RFP) is a magnetic confinement device that promises to meet these requirements with power densities comparable to those in existing fission power plants. To establish feasibility of such an RFP reactor, a practical design for a first wall capable of withstanding high levels of cyclic neutron wall loadings is needed. Associated with the neutron flux in the proposed RFP reactor is a time averaged heat flux of 4.5 MW/sq m with a conservatively estimated transient peak approximately twice the average value. The design for a modular first wall made from a high-strength copper alloy that will meet these requirements of cyclic thermal loading is presented. The heat removal from the wall is by subcooled water flowing in straight tubes at high linear velocities.

  4. Fuel Grading Study on a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor

    SciTech Connect

    Ilas, Germina; Primm, Trent

    2009-11-01

    An engineering design study that would enable the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium fuel is ongoing at Oak Ridge National Laboratory. The computational models used to search for a low-enriched uranium (LEU) fuel design that would meet the requirements for the conversion study, and the recent results obtained with these models during FY 2009, 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 high-enriched uranium fuel core. These studies 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.

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

    SciTech Connect

    Ambrosek, Richard Garry; Ingram, Frederick William

    1999-11-01

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

  7. Large break loss-of-coolant accident analyses for the high flux isotope reactor

    SciTech Connect

    Taleyarkhan, R.P. )

    1989-01-01

    The US Department of Energy's High Flux Isotope Reactor (HFIR) was analyzed to evaluate it's response to a spectrum of loss-of-coolant accidents (LOCAs) with potential for leading to core damage. The MELCOR severe accident analysis code (version 1.7.1) was used to evaluate the overall dynamic response of HFIR. Before conducting LOCA analyses, the steady-state thermal-hydraulic parameters evaluated by MELCOR for various loop sections were verified against steady-state operating data. Thereafter, HFIR depressurization tests were simulated to evaluate the system pressure change for a given depletion in coolant inventory. Interesting and important safety-related phenomena were observed. The current analyses (which should be considered preliminary) that occur over a period from 1 to 3 seconds do not lead to core wide fuel melting. Core fluid flashing during the initial rapid depressurization does cause fuel temperature excursions due to adiabatic-like heatup. 3 refs., 4 figs.

  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. 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.

  10. Hydrogen Explosion Analysis for Cold Source Installation at the High Flux Isotope Reactor

    SciTech Connect

    Cook, David Howard

    2008-01-01

    Installation of a cold neutron source in the High Flux Isotope Reactor (HFIR) involved introduction of pressurized, cryogenic hydrogen into the facility and created explosion hazards to reactor safety-related equipment and personnel. Evaluation of potential hydrogen releases and facility/personnel consequences as a result of explosions was a key part of the safety analyses submitted to the DOE to obtain approval for testing and operation with hydrogen. This paper involves a description of the various hydrogen release and explosion consequence analyses that were performed. The range of explosion calculations involved (1) a detonation analysis using a 2D-transient CTH code model, (2) various BLAST/FX code models to estimate structural damage from equivalent point TNT sources, (3) a BLASTX code model to propagate shock and gas flow overpressures from a point TNT source, (4) a spreadsheet that combined a TNT-quivalence model and strong deflagration methods, and (5) a hydrogen jet model to evaluate potential high pressure jet releases.

  11. The ignition physics study group supports the compact ignition tokamak and engineering test reactor programs

    SciTech Connect

    Sheffield, J.

    1987-01-01

    This report presents a collection of Vugraphs dealing with the Compact Ignition Tokamak (CIT) and the Engineering Test Reactor (ETR). The role of the Ignition Physics Study Group is defined. Several design goals are presented. (JDH)

  12. Power Distribution Analysis for the ORNL High Flux Isotope Reactor Critical Experiment 3

    SciTech Connect

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

    2010-01-01

    The mission of the Reduced Enrichment for Research and Test Reactors Program is to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors, as well as radioisotope production processes, to low-enriched uranium (LEU) fuel and targets. Oak Ridge National Laboratory (ORNL) is currently reviewing the design bases and key operating criteria including fuel operating parameters, enrichment-related safety analyses, fuel performance, and fuel fabrication in regard to converting the fuel of the High Flux Isotope Reactor (HFIR) from HEU to LEU. The purpose of this study is to validate Monte Carlo methods currently in use for conversion analyses. The methods have been validated for the prediction offlux values in the reactor target, reflector, and beam tubes, but this study focuses on the prediction of the power density profile in the core. Power distributions were calculated in the fuel elements of the HFIR, a research reactor at ORNL, via MCNP and were compared to experimentally obtained data. This study was performed to validate Monte Carlo methods for power density calculations and to observe biases. A current three-dimensional MCNP model was modified to replicate the 1965 HFIR Critical Experiment 3 (HFIRCE-3). In this experiment, the power profile was determined by counting the gamma activity at selected locations in the core. 'Foils' (chunks of fuel meat and clad) were punched out of the fuel elements in HFIRCE-3 following irradiation, and experimental relative power densities were obtained by measuring the activity of these foils and comparing each foil's activity to the activity of a normalizing foil. This analysis consisted of calculating corresponding activities by inserting volume tallies into the modified MCNP model to represent the punchings. The average fission density was calculated for each foil location and then normalized to the reference foil

  13. A Review of Proposed Upgrades to the High Flux Isotope Reactor and Potential Impacts to Reactor Vessel Integrity

    SciTech Connect

    Simonen, Fredric A.

    2001-05-31

    The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was scheduled in October 2000 to implement design upgrades that include the enlargement of the HB-2 and HB-4 beam tubes. Higher dose rates and higher radiation embrittlement rates were predicted for the two beam-tube nozzles and surrounding vessel areas. ORNL had performed calculations for the upgraded design to show that vessel integrity would be maintained at acceptable levels. Pacific Northwest National Laboratory (PNNL) was requested by the U.S. Department of Energy Headquarters (DOE/HQ) to perform an independent peer review of the ORNL evaluations. PNNL concluded that the calculated probabilities of failure for the HFIR vessel during hydrostatic tests and for operational conditions as estimated by ORNL are an acceptable basis for selecting pressures and test intervals for hydrostatic tests and for justifying continued operation of the vessel. While there were some uncertainties in the embrittlement predictions, the ongoing efforts at ORNL to measure fluence levels at critical locations of the vessel wall and to test materials from surveillance capsules should be effective in dealing with embrittlement uncertainties. It was recommended that ORNL continue to update their fracture mechanics calculations to reflect methods and data from ongoing research for commercial nuclear power plants. Such programs should provide improved data for vessel fracture mechanics calculations.

  14. Lessons Learned in the Update of a Safety Limit for the High Flux Isotope Reactor

    SciTech Connect

    Cook, David Howard

    2009-01-01

    A recent unreviewed safety question (USQ) regarding a portion of the High Flux Isotope Reactor (HFIR) transient decay heat removal analysis focused on applicability of a heat transfer correlation at the low flow end of reactor operations. During resolution of this issue, review of the correlations used to establish the safety limit (SL) on reactor flux-to-flow ratio revealed the need to change the magnitude of the SL at the low flow end of reactor operations and the need to update the hot spot fuel damage criteria to incorporate current knowledge involving parallel channel flow stability. Because of the original safety design strategy for the reactor, resolution of the issues for the flux-to-flow ratio involved reevaluation of all key process variable SLs and limiting control settings (LCSs) using the current version of the heat transfer analysis code for the reactor. Goals of the work involved updating and upgrading the SL analysis where necessary, while preserving the safety design strategy for the reactor. Changes made include revisions to the safety design criteria at low flows to address the USQ, update of the process- and analysis input-variable uncertainty considerations, and upgrade of the safety design criteria at high flow. The challenges faced during update/upgrade of this SL and LCS are typical of the problems found in the integration of safety into the design process for a complex facility. In particular, the problems addressed in the area of instrument uncertainties provide valuable lessons learned for establishment and configuration control of SLs for large facilities.

  15. 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. PMID:22739969

  16. 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.

  17. Simulation of the Performance of a Fundamental Neutron Physics Beamline at the High Flux Isotope Reactor.

    PubMed

    Mahurin, Rob; Greene, Geoffrey; Kohler, Paul; Cianciolo, Vince

    2005-01-01

    We study the expected performance of the proposed fundamental neutron physics beamline at the upgraded High Flux Isotope Reactor at Oak Ridge National Laboratory. A curved neutron guide transmits the neutrons from the new cold source into a guide hall. A novel feature of the proposed guide is the use of vertical focussing to increase the flux for experiments that require relatively small cross-section beams. We use the simulation code IB to model straight, multi-channel curved, and tapered guides of various m values. Guide performance for the current NPDGamma and proposed abBA experiments is evaluated. PMID:27308114

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

    SciTech Connect

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

    2009-05-03

    The High Flux Isotope Reactor resumed operation in June of 2007 with a super-critical hydrogen cold source in horizontal beam tube 4. Cold guide 4 is a guide system designed to deliver neutrons from this source at 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.

  19. Simulation of the Performance of a Fundamental Neutron Physics Beamline at the High Flux Isotope Reactor

    SciTech Connect

    Mahurin, R.; Greene, Geoffrey L; Koehler, Paul Edward; Cianciolo, Vince

    2005-05-01

    We study the expected performance of the proposed fundamental neutron physics beamline at the upgraded High Flux Isotope Reactor at Oak Ridge National Laboratory. A curved neutron guide transmits the neutrons from the new cold source into a guide hall. A novel feature of the proposed guide is the use of vertical focusing to increase the flux for experiments that require relatively small cross-section beams. We use the simulation code IB to model straight, multi-channel curved, and tapered guides of various m values. Guide performance for the current NPDGamma and proposed abBA experiments is evaluated.

  20. Simulation of the Performance of a Fundamental Neutron Physics Beamline at the High Flux Isotope Reactor

    PubMed Central

    Mahurin, Rob; Greene, Geoffrey; Kohler, Paul; Cianciolo, Vince

    2005-01-01

    We study the expected performance of the proposed fundamental neutron physics beamline at the upgraded High Flux Isotope Reactor at Oak Ridge National Laboratory. A curved neutron guide transmits the neutrons from the new cold source into a guide hall. A novel feature of the proposed guide is the use of vertical focussing to increase the flux for experiments that require relatively small cross-section beams. We use the simulation code IB to model straight, multi-channel curved, and tapered guides of various m values. Guide performance for the current NPDGamma and proposed abBA experiments is evaluated. PMID:27308114

  1. Neutron Radiography Facility at IBR-2 High Flux Pulsed Reactor: First Results

    NASA Astrophysics Data System (ADS)

    Kozlenko, D. P.; Kichanov, S. E.; Lukin, E. V.; Rutkauskas, A. V.; Bokuchava, G. D.; Savenko, B. N.; Pakhnevich, A. V.; Rozanov, A. Yu.

    A neutron radiography and tomography facilityhave been developed recently at the IBR-2 high flux pulsed reactor. The facility is operated with the CCD-camera based detector having maximal field of view of 20x20 cm, and the L/D ratio can be varied in the range 200 - 2000. The first results of the radiography and tomography experiments with industrial materials and products, paleontological and geophysical objects, meteorites, are presented.

  2. High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

    SciTech Connect

    Guillen, Donna; Greenwood, Lawrence R.; Parry, James

    2014-06-22

    A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated for up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.

  3. 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

  4. 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.

  5. 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.

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

    SciTech Connect

    Primm, Trent

    2008-01-01

    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. Such a model was built based on the available description parameters as provided by the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analysis performed with the model constructed was 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 making LEU fuel a safe alternative fuel for the reactor core.

  7. 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.

  8. 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.

  9. Radiation Dosimetry of the Pressure Vessel Internals of the High Flux Beam Reactor

    NASA Astrophysics Data System (ADS)

    Holden, Norman E.; Reciniello, Richard N.; Hu, Jih-Perng; Rorer, David C.

    2003-06-01

    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, 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™ 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 rates, the Monte Carlo MCNP code and geometric progressive MicroShield code were used to model the gamma-ray transport and dose buildup.

  10. Optimization of Depletion Modeling and Simulation for the High Flux Isotope Reactor

    SciTech Connect

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

    2015-01-01

    Monte Carlo based depletion tools used for the high-fidelity modeling and simulation of the High Flux Isotope Reactor (HFIR) come at a great computational cost; finding sufficient approximations is necessary to make the use of these tools feasible. The optimization of the neutronics and depletion model for the HFIR is based on two factors: (i) the explicit representation of the involute fuel plates with sets of polyhedra and (ii) the treatment of depletion mixtures and control element position during depletion calculations. A very fine representation (i.e., more polyhedra in the involute plate approximation) does not significantly improve simulation accuracy. The recommended representation closely represents the physical plates and ensures sufficient fidelity in regions with high flux gradients. Including the fissile targets in the central flux trap of the reactor as depletion mixtures has the greatest effect on the calculated cycle length, while localized effects (e.g., the burnup of specific isotopes or the power distribution evolution over the cycle) are more noticeable consequences of including a critical control element search or depleting burnable absorbers outside the fuel region.

  11. 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.

  12. 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.

  13. Seismic hazard studies for the High Flux Beam Reactor at Brookhaven National Laboratory

    SciTech Connect

    Costantino, C.J.; Heymsfield, E. . Dept. of Civil Engineering); Park, Y.J.; Hofmayer, C.H. )

    1991-01-01

    This paper presents the results of a calculation to determine the site specific seismic hazard appropriate for the deep soil site at Brookhaven National Laboratory (BNL) which is to be used in the risk assessment studies being conducted for the High Flux Beam Reactor (HFBR). The calculations use as input the seismic hazard defined for the bedrock outcrop by a study conducted at Lawrence Livermore National Laboratory (LLNL). Variability in site soil properties were included in the calculations to obtain the seismic hazard at the ground surface and compare these results with those using the generic amplification factors from the LLNL study. 9 refs., 8 figs.

  14. PREPARING THE HIGH FLUX ISOTOPE REACTOR FOR CONVERSION TO LOW ENRICHED URANIUM FUEL ? RETURN TO 100 MW

    SciTech Connect

    Smith, Kevin Arthur; Primm, Trent

    2009-01-01

    The feasibility of low-enriched uranium (LEU) fuel as a replacement for the current, high enriched uranium (HEU) fuel for the High Flux Isotope Reactor (HFIR) has been under study since 2006. Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting to LEU fuel requires returning the reactor power to 100 MW from 85 MW. The analyses required to up-rate the reactor power and the methods to perform these analyses are discussed. Comments regarding the regulatory approval process are provided along with a conceptual schedule.

  15. 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.

  16. 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.

  17. 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.

  18. Modeling of divertor geometry effects in China fusion engineering testing reactor by SOLPS/B2-Eirene

    SciTech Connect

    Zhao, M. L.; Chen, Y. P.; Li, G. Q.; Luo, Z. P.; Guo, H. Y.; Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031; General Atomics, P.O. Box 85608, San Diego, California 92186 ; Ye, M. Y.; Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 ; Tendler, M.

    2014-05-15

    The China Fusion Engineering Testing Reactor (CFETR) is currently under design. The SOLPS/B2-Eirene code package is utilized for the design and optimization of the divertor geometry for CFETR. Detailed modeling is carried out for an ITER-like divertor configuration and one with relatively open inner divertor structure, to assess, in particular, peak power loading on the divertor target, which is a key issue for the operation of a next-step fusion machine, such as ITER and CFETR. As expected, the divertor peak heat flux greatly exceeds the maximum steady-state heat load of 10 MW/m{sup 2}, which is a limit dictated by engineering, for both divertor configurations with a wide range of edge plasma conditions. Ar puffing is effective at reducing divertor peak heat fluxes below 10 MW/m{sup 2} even at relatively low densities for both cases, favoring the divertor configuration with more open inner divertor structure.

  19. Hypersonic engine component experiments in high heat flux, supersonic flow environment

    NASA Technical Reports Server (NTRS)

    Gladden, Herbert J.; Melis, Matthew E.

    1993-01-01

    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. Even though progress has been made in the computational understanding of fluid dynamics and the physics/chemistry of high speed flight, there is also a need for experimental facilities capable of providing a high heat flux environment for testing component concepts and verifying/calibrating these analyses. A hydrogen/oxygen rocket engine heat source was developed at the NASA Lewis Research Center as one element in a series of facilities at national laboratories designed to fulfill this need. This 'Hot Gas Facility' is capable of providing heat fluxes up to 450 w/sq cm on flat surfaces and up to 5,000 w/sq cm at the leading edge stagnation point of a strut in a supersonic flow stream. Gas temperatures up to 3050 K can also be attained. Two recent experimental programs conducted in this facility are discussed. The objective of the first experiment 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. Macrophotographic data from a coated leading edge model show progressive degradation over several thermal cycles at aerothermal conditions representative of high Mach number flight. The objective of the second experiment is to assess the capability of cooling a porous surface exposed to a high temperature, high velocity flow environment and to provide a heat transfer data base for a design procedure. Experimental results from transpiration cooled surfaces in a supersonic flow environment are presented.

  20. 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.

  1. 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.

  2. Engine tests using high-sulfur diesel fuel. Final report

    SciTech Connect

    Frame, E.A.; Moon, R.B.

    1980-09-01

    This report covers the engine test evaluation of an organo-zinc additive for its effectiveness in combating the deleterious effects of using high-sulfur diesel fuel in a two-cycle U.S. Army diesel engine. The report also covers the 6V-53T testing of a preservative engine oil which in previous testing had shown promise in controlling the effects of using high-sulfur fuel.

  3. Upgrade of the wide-angle neutron diffractometer at the high flux isotope reactor

    SciTech Connect

    Katano, S.; Morii, Y.; Child, H.R.

    1997-07-15

    The Wide-Angle Neutron Diffractometer (WAND) is a flat-cone geometry diffractometer located at the High Flux Reactor (HFIR). This instrument is currently being upgraded. The central part of this upgrade is the development of a new curved one-dimensional position sensitive detector which covers a 125 degree angular range with an effective radius of 71 cm. This detector will be a multi-anode (624 anodes on a 0.2 degree pitch) {sup 3}He gas-filled proportional counter. This totally new system will give high resolution, good uniformity and high counting range - a maximum capability of 10{sup 5} cps/pixel and a 10{sup 7} cps overall. A prototype of this detector has shown that these design targets can be met. The new WAND will greatly broaden its capabilities for single-crystal diffraction experiments and for time-resolved measurements.

  4. Transient hot-spot analysis of the high-flux isotope reactor

    SciTech Connect

    Sofu, T.; Dodds, H.L. ); Cook, D.H. )

    1992-01-01

    Consideration of fuel element failure and the possibility of failure propagation is part of the overall safety approach for the Oak Ridge National Laboratory's High-Flux Isoptope Reactor (FHIR). There is a need to establish a technical basis for understanding the initiators that lead to fuel melting, the phenomena of fuel melting, and its consequences in order to reevaluate core damage criteria and to ensure that the HFIR core is designed to avoid or mitigate such accidents to the greatest extent possible. In this work, a computer model for the transient analysis of the HFIR is developed and applied for the study of hot-spot behavior during rapid HFIR transients. An earlier model that was developed for the analysis of reactivity transients of U-Al-dispersion-fueled research reactors is used as the base model. It includes point reactor kinetics for neutronics, a nonhomogeneous, nonequilibrium representation of the two-phase thermal hydraulics, and one-dimensional, spatially averaged heat conduction through fuel plates. The flow and heat transfer regimes considered are single-phase liquid flow, subcooled boiling, bulk boiling, film boiling, and single-phase vapor flow.

  5. Validation of KENO V.a Code for High Flux Isotope Reactor (HFIR)

    SciTech Connect

    Primm, Trent

    2009-01-01

    The core of the High Flux Isotope Reactor (HFIR) is composed of two concentric annular elements, inner and outer, each containing highly enriched uranium fuel as a mixture of triuranium octoxide (U3O8) and aluminum encapsulated within aluminum alloy plates. The fuel plates are of involute shape and the fuel within the plates has a distribution across the plate width. Previous KENO code validation efforts have used a relatively simple single region homogeneous fuel model for each of the two annular regions by assuming that the materials in each were homogenized within the total volume of the fueled region. The computed results have tended to be about 2 to 3% greater than experimentally measured results. To improve computed results, a multi-zone fuel model was developed and used to validate the KENO code.

  6. 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.

  7. 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.

  8. Advanced high temperature heat flux sensors

    NASA Technical Reports Server (NTRS)

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

    1983-01-01

    To fully characterize advanced high temperature heat flux sensors, calibration and testing is required at full engine temperature. This required the development of unique high temperature heat flux test facilities. These facilities were developed, are in place, and are being used for advanced heat flux sensor development.

  9. Design optimization and experimental testing of the High-Flux Test Module of IFMIF

    NASA Astrophysics Data System (ADS)

    Leichtle, D.; Arbeiter, F.; Dolensky, B.; Fischer, U.; Gordeev, S.; Heinzel, V.; Ihli, T.; Moeslang, A.; Simakov, S. P.; Slobodchuk, V.; Stratmanns, E.

    2009-04-01

    The design of the High-Flux Test Module of the International Fusion Material Irradiation Facility has been developed continuously in the past few years. The present paper highlights recent design achievements, including a thorough state-of-the-art validation assessment of CFD tools and models. Along with design related analyses exercises on manufacturing procedures have been performed. Recommendations for the use of container, rig, and capsule materials as well as recent progress in brazing of electrical heaters are discussed. A test matrix starting from High-Flux Test Module compartments, i.e. segments of the full module, with heated dummy rigs up to the full-scale module with instrumented irradiation rigs has been developed and the appropriate helium gas loop has been designed conceptually. A roadmap of the envisaged experimental activities is presented in accordance with the test loop facility construction and mock-up design and fabrication schedules.

  10. 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.

  11. 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. PMID:15220719

  12. 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).

  13. Reactor Physics Studies of Reduced-Tantaulum-Content Control and Safety Elements for the High Flux Isotope Reactor

    SciTech Connect

    Primm, R.T., III

    2003-11-01

    Some of the unirradiated High Flux Isotope Reactor (HFIR) control elements discharged during the late 1990s were observed to have cladding damage--local swelling or blistering. The cladding damage was limited to the tantalum/europium interface of the element and is thought to result from interaction of hydrogen and europium to form a compound of lower density than europium oxide, thus leading to a ''blistering'' of the control plate cladding. Reducing the tantalum loading in the control plates should help preclude this phenomena. The impact of the change to the control plates on the operation of the reactor was assessed. Regarding nominal, steady-state reactor operation, the impact of the change in the power distribution in the core due to reduced tantalum content was calculated and found to be insignificant. The magnitude and impact of the change in differential control element worth was calculated, and the differential worths of reduced tantalum elements vs the current elements from equivalent-burnup critical configurations were determined to be unchanged within the accuracy of the computational method and relevant experimental measurements. The location of the critical control elements symmetric positions for reduced tantalum elements was found to be 1/3 in. less withdrawn relative to existing control elements regardless of the value of fuel cycle burnup (time in the fuel cycle). The magnitude and impact of the change in the shutdown margin (integral rod worth) was assessed and found to be unchanged. Differential safety element worth values for the reduced-tantalum-content elements were calculated for postulated accident conditions and were found to be greater than values currently assumed in HFIR safety analyses.

  14. 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.

  15. A system analysis computer model for the High Flux Isotope Reactor (HFIRSYS Version 1)

    SciTech Connect

    Sozer, M.C.

    1992-04-01

    A system transient analysis computer model (HFIRSYS) has been developed for analysis of small break loss of coolant accidents (LOCA) and operational transients. The computer model is based on the Advanced Continuous Simulation Language (ACSL) that produces the FORTRAN code automatically and that provides integration routines such as the Gear`s stiff algorithm as well as enabling users with numerous practical tools for generating Eigen values, and providing debug outputs and graphics capabilities, etc. The HFIRSYS computer code is structured in the form of the Modular Modeling System (MMS) code. Component modules from MMS and in-house developed modules were both used to configure HFIRSYS. A description of the High Flux Isotope Reactor, theoretical bases for the modeled components of the system, and the verification and validation efforts are reported. The computer model performs satisfactorily including cases in which effects of structural elasticity on the system pressure is significant; however, its capabilities are limited to single phase flow. Because of the modular structure, the new component models from the Modular Modeling System can easily be added to HFIRSYS for analyzing their effects on system`s behavior. The computer model is a versatile tool for studying various system transients. The intent of this report is not to be a users manual, but to provide theoretical bases and basic information about the computer model and the reactor.

  16. Breeder Reprocessing Engineering Test

    SciTech Connect

    Burgess, C.A.; Meacham, S.A.

    1984-01-01

    The Breeder Reprocessing Engineering Test (BRET) is a developmental activity of the US Department of Energy to demonstrate breeder fuel reprocessing technology while closing the fuel cycle for the Fast Flux Test Facility (FFTF). It will be installed in the existing Fuels and Materials Examination Facility (FMEF) at the Hanford Site near Richland, Washington, The major objectives of BRET are: (1) close the US breeder fuel cycle; (2) develop and demonstrate reprocessing technology and systems for breeder fuel; (3) provide an integrated test of breeder reactor fuel cycle technology - rprocessing, safeguards, and waste management. BRET is a joint effort between the Westinghouse Hanford Company and Oak Ridge National Laboratory. 3 references, 2 figures.

  17. 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.

  18. 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

  19. 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

  20. SSME improved characterization using highly instrumented engine test data

    NASA Astrophysics Data System (ADS)

    Piekarski, Brian; Leahy, Joseph C.

    1992-07-01

    This paper discusses the analysis of Space Shuttle Main Engine (SSME) 3001 test data and its applications to steady-state models. Engine 3001 is a highly instrumented Phase II (current configuration) SSME that was tested at the Technology Test Bed facility located at Marshall Space Flight Center. A test series covering a carefully selected range of operating conditions was performed during 1991. An analysis of pressure, temperature and flowrate data was performed to evaluate the ability of steady-state models to accurately characterize the internal system of an SSME. Much of the special instrumentation was used to acquire data never before measured at the engine level prior to Engine 3001 testing. The analysis revealed that modifications are needed to improve model calculations of internal Hydrogen and Oxygen flows. Modifying steady-state models to more accurately simulate the internal flow system of an SSME should improve both engine performance evaluations and prediction capabilities.

  1. SSME improved characterization using highly instrumented engine test data

    NASA Technical Reports Server (NTRS)

    Piekarski, Brian; Leahy, Joseph C.

    1992-01-01

    This paper discusses the analysis of Space Shuttle Main Engine (SSME) 3001 test data and its applications to steady-state models. Engine 3001 is a highly instrumented Phase II (current configuration) SSME that was tested at the Technology Test Bed facility located at Marshall Space Flight Center. A test series covering a carefully selected range of operating conditions was performed during 1991. An analysis of pressure, temperature and flowrate data was performed to evaluate the ability of steady-state models to accurately characterize the internal system of an SSME. Much of the special instrumentation was used to acquire data never before measured at the engine level prior to Engine 3001 testing. The analysis revealed that modifications are needed to improve model calculations of internal Hydrogen and Oxygen flows. Modifying steady-state models to more accurately simulate the internal flow system of an SSME should improve both engine performance evaluations and prediction capabilities.

  2. FFTF (Fast Flux Test Facility) Reactor Characterization Program: Absolute Fission-rate Measurements

    SciTech Connect

    Fuller, J.L.; Gilliam, D.M.; Grundl, J.A.; Rawlins, J.A.; Daughtry, J.W.

    1981-05-01

    Absolute fission rate measurements using modified National Bureau of Standards fission chambers were performed in the Fast Flux Test Facility at two core locations for isotopic deposits of {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu. Monitor chamber results at a third location were analyzed to support other experiments involving passive dosimeter fission rate determinations.

  3. FFTF (FAST FLUX TEST FACILITY) REACTOR CHARACTERIZATION PROGRAM ABSOLUTE FISSION RATE MEASUREMENTS

    SciTech Connect

    FULLER JL; GILLIAM DM; GRUNDL JA; RAWLINS JA; DAUGHTRY JW

    1981-05-01

    Absolute fission rate measurements using modified National Bureau of Standards fission chambers were performed in the Fast Flux Test Facility at two core locations for isotopic deposits of {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu. Monitor chamber results at a third location were analyzed to support other experiments involving passive dosimeter fission rate determinations.

  4. 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.

  5. 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.

  6. 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

  7. 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.

  8. 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

  9. 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.

  10. Extraction of gadolinium from high flux isotope reactor control plates. [Alternative method

    SciTech Connect

    Kohring, M.W.

    1987-04-01

    Gadolinium-153 is an important radioisotope used in the diagnosis of various bone disorders. Recent medical and technical developments in the detection and cure of osteoporosis, a bone disease affecting an estimated 50 million people, have greatly increased the demand for this isotope. The Oak Ridge National Laboratory (ORNL) has produced /sup 153/Gd since 1980 primarily through the irradiation of a natural europium-oxide powder followed by the chemical separation of the gadolinium fraction from the europium material. Due to the higher demand for /sup 153/Gd, an alternative production method to supplement this process has been investigated. This process involves the extraction of gadolinium from the europium-bearing region of highly radioactive, spent control plates used at the High Flux Isotope Reactor (HFIR) with a subsequent re-irradiation of the extracted material for the production of the /sup 153/Gd. Based on the results of experimental and calculational analyses, up to 25 grams of valuable gadolinium (greater than or equal to60% enriched in /sup 152/Gd) resides in the europium-bearing region of the HFIR control components of which 70% is recoverable. At a specific activity yield of 40 curies of /sup 153/Gd for each gram of gadolinium re-irradiated, 700 one-curie sources can be produced from each control plate assayed.

  11. 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.

  12. 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.

  13. 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.

  14. 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

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

    NASA Astrophysics Data System (ADS)

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

    1992-07-01

    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.

  16. 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.

  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. 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)

  19. A Performance-Based Training Qualification Guide/Checklist Developed for Reactor Operators at the High Flux Beam Reactor at Brookhaven National Laboratory.

    ERIC Educational Resources Information Center

    McNair, Robert C.

    A Performance-Based Training (PBT) Qualification Guide/Checklist was developed that would enable a trainee to attain the skills, knowledge, and attitude required to operate the High Flux Beam Reactor at Brookhaven National Laboratory. Design of this guide/checklist was based on the Instructional System Design Model. The needs analysis identified…

  20. Scientific Upgrades at the Oak Ridge National Laboratory High Flux Isotope Reactor

    SciTech Connect

    Selby, Douglas L; Jones, Amy; Crow, Lowell

    2012-01-01

    The United States Department of Energy is sponsoring a number of projects that will provide scientific upgrades to the neutron science facilities associated with the High Flux Isotope Reactor (HFIR) located at Oak Ridge National Laboratory. Funding for the first upgrade project was initiated in 1996 and all presently identified upgrade projects are expected to be completed by the end of 2003. The upgrade projects include: (1) larger beam tubes, (2) a new monochromator drum for the HB-1 beam line, (3) a new HB-2 beam line system that includes one thermal guide and a new monochromator drum, (4) new instruments for the HB-2 beamline, (5) a new monochromator drum for the HB-3 beam line, (6) a supercritical hydrogen cold source system to be retrofitted into the HB-4 beam tube, (7) a 3.5 kW refrigeration system at 20 K to support the cold source and a new building to house it, (8) a new HB-4 beam line system composed of four cold neutron guides with various mirror coatings and associated shielding, (9) a number of new instruments for the cold beams including two new SANS instruments, and (10) construction of support buildings. This paper provides a short summary of these projects including their present status and schedule.

  1. Scientific Upgrades at the High Flux Isotope Reactor at Oak Ridge National Laboratory

    SciTech Connect

    Selby, Douglas L; Smith, Gregory Scott

    2010-01-01

    The United States Department of Energy is sponsoring a number of projects that will provide scientific upgrades to the neutron science facilities associated with the high Flux Isotope Reactor (HFIR) located at Oak Ridge National Laboratory. Funding for the first upgrade project was initiated in 1996 and all presently identified upgrade projects are expected to be completed by the end of 2003. The upgrade projects include: (1) larger beam tubes, (2) a new monochromator drum for the HB-1 beam line, (3) a new HB-2 beam line system that includes one thermal guide and a new monochromator drum, (4) new instruments for the HB-2 beamline, (5) a new monochromator drum for the HB-3 beam line, (6) a supercritical hydrogen cold source system to be retrofitted into the HB-4 beam tube, (7) a 3.5 kW refrigeration system at 20 K to support the cold source and a new building to house it, (8) a new HB-4 beam line system composed of four cold neutron guides with various mirror coatings and associated shielding, (9) a number of new instruments for the cold beams including two new SANS instruments, and (10) construction of support buildings. This paper provides a short summary of these projects including their present status and schedule.

  2. 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.

  3. 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.

  4. 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

  5. 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

  6. 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.

  7. 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.

  8. Evaporation Basin Test Reactor Area, Idaho National Engineering Laboratory: Environmental assessment

    SciTech Connect

    Not Available

    1991-12-01

    The Department of Energy (DOE) has prepared an environmental assessment (EA), DOE/EA-0501, on the construction and operation of the proposed Evaporation Basin at the Test Reactor Area (TRA) at the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho. Based on the analyses in the EA, DOE has determined that the proposed action is not a major Federal action significantly affecting the quality of the human environment, within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, the preparation of an environmental impact statement (EIS) is not required, and the Department is issuing this Finding of No Significant Impact.

  9. Flux and spectrum of reactor antineutrinos

    NASA Astrophysics Data System (ADS)

    Kopeikin, V. I.

    2012-02-01

    In order to perform reactor experiments aimed at studying the nature of the neutrino and measurements in the realms of geo- and astrophysical neutrinos and to meet practical requirements in this field, it is highly desirable to obtain deeper insight into the operation of nuclear reactors as a source of antineutrinos. The fluxes and spectra of neutrinos from a reactor in the on and off modes and from a reservoir intended for storing a spent reactor fuel and situated near the reactor being considered are calculated. Features that are peculiar to the flux and spectrum of reactor antineutrinos and which are of importance for implementing and interpreting experiments, but which were disregarded previously, are analyzed here.

  10. 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).

  11. 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

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

    NASA Astrophysics Data System (ADS)

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

    2014-12-01

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

  13. 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].

  14. 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.

  15. 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.

  16. 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

  17. Reference (Axially Graded) Low Enriched Uranium Fuel Design for the High Flux Isotope Reactor (HFIR)

    SciTech Connect

    Ilas, Germina; Primm, Trent

    2010-01-01

    During the past five years, staff at the Oak Ridge National Laboratory (ORNL) have studied the issue of whether the HFIR could be converted to low enriched uranium (LEU) fuel without degrading the performance of the reactor. Using state-of-the-art reactor physics methods and behind-the-state-of-the-art thermal hydraulics methods, the staff have developed fuel plate designs (HFIR uses two types of fuel plates) that are believed to meet physics and thermal hydraulic criteria provided the reactor power is increased from 85 to 100 MW. The paper will present a defense of the results by explaining the design and validation process. A discussion of the requirements for showing applicability of analyses to approval for loading the fuel to HFIR lead test core irradiation currently scheduled for 2016 will be provided. Finally, the potential benefits of upgrading thermal hydraulics methods will be discussed.

  18. 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.

  19. 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

  20. High heat flux testing of 12-14Cr ODS ferritic steels

    NASA Astrophysics Data System (ADS)

    Pintsuk, G.; Oksiuta, Z.; Linke, J.; Baluc, N.

    2010-01-01

    The thermal performance of Fe-(12-14)Cr-2W-0.3Ti-0.3Y 2O 3 ODS reduced activation ferritic steels, which are considered as candidate first wall materials for the future fusion power reactors and were manufactured by mechanical alloying in hydrogen and hot isostatic pressing, was assessed by high heat flux (HHF) testing with the electron beam JUDITH facility at the Forschungszentrum Jülich (FZJ), Germany. An analysis of the microhardness and microstructure of the specimens was done before and after HHF tests. In general, both materials present a ferritic (α-Fe, bcc) microstructure with a wide range of grain sizes from 100 to 500 nm up to a few micrometers. The coarse grains are almost dislocation-free, while the smaller ones are surrounded by tangles of dislocations. Oxide and carbide impurities (about a few hundreds nm in size) and a high density of Y-Ti-O nano-clusters, with a mean size of about 5 nm, are also present. The microhardness, density and tensile strength of the 14Cr material are slightly larger than those of the 12Cr material. HHF tests revealed that there is no difference in thermal performance, level of degradation and erosion behaviour of 12Cr and 14Cr ODS steels. The onset of melting of the materials occurs for an energy density between 1 and 1.5 MJ/m 2. Below this value only some kind of thermal etching takes place. This is a significant improvement compared to stainless steel, for which severe plastic deformation at the material surface was observed.

  1. Plasma engineering models of tandem mirror devices with high-field test-cell inserts

    SciTech Connect

    Fenstermacher, M.E.; Campbell, R.B.

    1985-04-03

    Plasma physics and engineering models of tandem mirror devices operated with a high-field technology test-cell insert in the central cell, which have been incorporated recently in the TMRBAR tandem mirror reactor physics code, are described. The models include particle and energy balance in the test-cell region as well as the interactions between the test-cell particles and those flowing through the entire device. The code calculations yield consistent operating parameters for the test-cell, central cell, and end cell systems. A benchmark case for the MFTF-..cap alpha..+T configuration is presented which shows good agreement between the code results and previous calculations.

  2. Test experience, 490 N high performance (321 sec Isp) engine

    NASA Technical Reports Server (NTRS)

    Schoenman, L.; Rosenberg, S. D.; Jassowski, D. M.

    1992-01-01

    Engines with area ratios of 44:1 and 286:1 are tested by means of hot fire tests using the NTO/MMH bipropellant to maximize the performance of the combined technologies. The low-thrust engine systems are designed with oxidation resistant materials that can operate at temperatures of more than 2204 C for tens of hours. The chamber is attached to the injector in a configuration that prevents overheating of the injector, valve, and the spacecraft interface. Three injectors with 44:1 area ratios are capable of nominal specific impulse values of 309 sec, and a performance of 321 lbf-sec/lbm is noted for an all-welded engine assembly with area ratio of 286:1. The all-welded engine is shown to have an acceptable design margin for thermal characteristics. High-performance liquid apogee engines are shown to perform optimally when based on iridium/rhenium chamber technology, use of a special platelet injector, and the minimization of losses due to fuel-film cooling.

  3. 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.

  4. 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.

  5. 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)

  6. 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.

  7. 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.

  8. High heat flux test of a HIP-bonded first wall panel of reduced activation ferritic steel F-82H

    NASA Astrophysics Data System (ADS)

    Hatano, T.; Suzuki, S.; Yokoyama, K.; Kuroda, T.; Enoeda, M.

    2000-12-01

    Reduced activation ferritic steel F-82H is a primary candidate structural material of DEMO fusion reactors. In fabrication technology, development of the DEMO blanket in JAERI, a hot isostatic pressing (HIP) bonding method, especially for the first wall structure with built-in cooling tubes has been proposed. A HIP-bonded F-82H first wall panel was successfully fabricated with selected manufacturing parameters. A high heat flux test of the HIP-bonded F-82H first wall panel has been performed to examine the thermo-mechanical performance of the panel including the integrity of the HIP-bonded interfaces and the fatigue behavior. A maximum heat flux of 2.7 MW/m2 was applied to accelerate the fatigue test up to 5000 cycles in test blanket inserted ITER. The maximum temperature of the panel was ∼450°C under this heat flux. Through this test campaign, no damage such as cracks was observed on the surface of the panel, and no degradation in heat removal performance was observed either from the temperature responses. The thermal fatigue lifetime of the panel was found to be longer than the fatigue data obtained by mechanical testing.

  9. Temporal behavior of neutral particle fluxes in TFTR (Tokamak Fusion Test Reactor) neutral beam injectors

    SciTech Connect

    Kamperschroer, J.H.; Gammel, G.M.; Roquemore, A.L.; Grisham, L.R.; Kugel, H.W.; Medley, S.S.; O'Connor, T.E.; Stevenson, T.N.; von Halle, A.; Williams, M.D.

    1989-09-01

    Data from an E {parallel} B charge exchange neutral analyzer (CENA), which views down the axis of a neutral beamline through an aperture in the target chamber calorimeter of the TFTR neutral beam test facility, exhibit two curious effects. First, there is a turn-on transient lasting tens of milliseconds having a magnitude up to three times that of the steady-state level. Second, there is a 720 Hz, up to 20% peak-to-peak fluctuation persisting the entire pulse duration. The turn-on transient occurs as the neutralizer/ion source system reaches a new pressure equilibrium following the effective ion source gas throughput reduction by particle removal as ion beam. Widths of the transient are a function of the gas throughput into the ion source, decreasing as the gas supply rate is reduced. Heating of the neutalizer gas by the beam is assumed responsible, with gas temperature increasing as gas supply rate is decreased. At low gas supply rates, the transient is primarliy due to dynamic changes in the neutralizer line density and/or beam species composition. Light emission from the drift duct corroborate the CENA data. At high gas supply rates, dynamic changes in component divergence and/or spatial profiles of the source plasma are necessary to explain the observations. The 720 Hz fluctuation is attributed to a 3% peak-to-peak ripple of 720 Hz on the arc power supply amplified by the quadratic relationship between beam divergence and beam current. Tight collimation by CENA apertures cause it to accept a very small part of the ion source's velocity space, producing a signal linearly proportional to beam divergence. Estimated fluctuations in the peak power density delivered to the plasma under these conditions are a modest 3--8% peak to peak. The efffects of both phenomena on the injected neutral beam can be ameliorated by careful operion of the ion sources. 21 refs., 11 figs., 2 tabs.

  10. 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.

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

    NASA Astrophysics Data System (ADS)

    Koelbl, Mary E.

    1993-06-01

    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.

  12. 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.

  13. 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.

  14. 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

  15. Selection of support structure materials for irradiation experiments in the HFIR (High Flux Isotope Reactor) at temperatures up to 500 degrees C

    SciTech Connect

    Farrell, K.; Longest, A.W.

    1990-01-01

    The key factor in the design of capsules for irradiation of test specimens in the High Flux Isotope Reactor at preselected temperatures up to 500{degree}C utilizing nuclear heating is a narrow gas-filled gap which surrounds the specimens and controls the transfer of heat from the specimens through the wall of a containment tube to the reactor cooling water. Maintenance of this gap to close tolerances is dependent on the characteristics of the materials used to support the specimens and isolate them from the water. These support structure materials must have low nuclear heating rates, high thermal conductivities, and good dimensional stabilities under irradiation. These conditions are satisfied by certain aluminum alloys. One of these alloys, a powder metallurgy product containing a fine dispersion of aluminum oxide, is no longer manufactured. A new alloys of this type, with the trade name DISPAL, is determined to be a suitable substitute. 23 refs., 13 figs., 3 tabs.

  16. 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

  17. High-heat-flux testing of helium-cooled heat exchangers for fusion applications

    SciTech Connect

    Youchison, D.L.; Izenson, M.G.; Baxi, C.B.; Rosenfeld, J.H.

    1996-07-01

    High-heat-flux experiments on three types of helium-cooled divertor mock-ups were performed on the 30-kW electron beam test system and its associated helium flow loop at Sandia National Laboratories. A dispersion-strengthened copper alloy (DSCu) was used in the manufacture of all the mock-ups. The first heat exchanger provides for enhanced heat transfer at relatively low flow rates and much reduced pumping requirements. The Creare sample was tested to a maximum absorbed heat flux of 5.8 MW/m{sup 2}. The second used low pressure drops and high mass flow rates to achieve good heat removal. The GA specimen was tested to a maximum absorbed heat flux of 9 MW/m{sup 2} while maintaining a surface temperature below 400{degree}C. A second experiment resulted in a maximum absorbed heat flux of 34 MW/m{sup 2} and surface temperatures near 533{degree}C. The third specimen was a DSCu, axial flow, helium-cooled divertor mock-up filled with a porous metal wick which effectively increases the available heat transfer area. Low mass flow and high pressure drop operation at 4.0 MPa were characteristic of this divertor module. It survived a maximum absorbed heat flux of 16 MW/m{sup 2} and reached a surface temperature of 740{degree}C. Thermacore also manufactured a follow-on, dual channel porous metal-type heat exchanger, which survived a maximum absorbed heat flux of 14 MW/m{sup 2} and reached a maximum surface temperature of 690{degree}C. 11refs., 20 figs., 3 tabs.

  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. 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.

  20. High-flux testing of heat pipes for point-focus solar collector systems

    NASA Astrophysics Data System (ADS)

    Adkins, D. R.

    The Department of Energy is currently funding the development of a power conversion system that uses a parabolic-dish collector to direct solar energy onto a 25-kWe Stirling engine/generator system. As part of this program, heat-pipe solar receivers are being developed to transfer 65 kW of thermal energy from the focus of the concentrator to the heater-tubes of the engine. These receivers operate at temperatures up to 800 C and with peak flux loads approaching 100 W/cm(sup 2). Designing a wick structure that can withstand these operating conditions has been a major part of the heat-pipe receiver development activities. Recently, bench-scale tests were performed on two advanced artery structures that are candidates for full-scale receivers. The arteries were able to lift sodium over 40 cm to cool a 30-cm(sup 2) patch on the bench-scale receiver's surface. Peak flux levels of approximately 80 W/cm(sup 2) were applied to the receiver surface.

  1. High flux isotope reactor. Quarterly report, April, May, and June 1981

    SciTech Connect

    Corbett, B.L.; Poteet, K.H.

    1982-01-01

    Routine reactor operation with four end-of-cycle shutdowns resulted in an on-stream time of 92.3% for the quarter. The outer control pates were changed and a semi-annual core component inspection made.

  2. 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.

  3. High-Lift Engine Aeroacoustics Technology (HEAT) Test Program Overview

    NASA Technical Reports Server (NTRS)

    Zuniga, Fanny A.; Smith, Brian E.

    1999-01-01

    The NASA High-Speed Research program developed the High-Lift Engine Aeroacoustics Technology (HEAT) program to demonstrate satisfactory interaction between the jet noise suppressor and high-lift system of a High-Speed Civil Transport (HSCT) configuration at takeoff, climb, approach and landing conditions. One scheme for reducing jet exhaust noise generated by an HSCT is the use of a mixer-ejector system which would entrain large quantities of ambient air into the nozzle exhaust flow through secondary inlets in order to cool and slow the jet exhaust before it exits the nozzle. The effectiveness of such a noise suppression device must be evaluated in the presence of an HSCT wing high-lift system before definitive assessments can be made concerning its acoustic performance. In addition, these noise suppressors must provide the required acoustic attenuation while not degrading the thrust efficiency of the propulsion system or the aerodynamic performance of the high-lift devices on the wing. Therefore, the main objective of the HEAT program is to demonstrate these technologies and understand their interactions on a large-scale HSCT model. The HEAT program is a collaborative effort between NASA-Ames, Boeing Commercial Airplane Group, Douglas Aircraft Corp., Lockheed-Georgia, General Electric and NASA - Lewis. The suppressor nozzles used in the tests were Generation 1 2-D mixer-ejector nozzles made by General Electric. The model used was a 13.5%-scale semi-span model of a Boeing Reference H configuration.

  4. 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.

  5. 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%).

  6. Impact induced response spectrum for the safety evaluation of the high flux isotope reactor

    SciTech Connect

    Chang, S.J.

    1997-05-01

    The dynamic impact to the nearby HFIR reactor vessel caused by heavy load drop is analyzed. The impact calculation is carried out by applying the ABAQUS computer code. An impact-induced response spectrum is constructed in order to evaluate whether the HFIR vessel and the shutdown mechanism may be disabled. For the frequency range less than 10 Hz, the maximum spectral velocity of impact is approximately equal to that of the HFIR seismic design-basis spectrum. For the frequency range greater than 10 Hz, the impact-induced response spectrum is shown to cause no effect to the control rod and the shutdown mechanism. An earlier seismic safety assessment for the HFIR control and shutdown mechanism was made by EQE. Based on EQE modal solution that is combined with the impact-induced spectrum, it is concluded that the impact will not cause any damage to the shutdown mechanism, even while the reactor is in operation. The present method suggests a general approach for evaluating the impact induced damage to the reactor by applying the existing finite element modal solution that has been carried out for the seismic evaluation of the reactor.

  7. High-flux source of fusion neutrons for material and component testing

    SciTech Connect

    Baldwin, D. E.; Hooper, E. B.; Ryutov, D. D.; Thomassen, K. I.

    1999-01-07

    The inner part of a fusion reactor will have to operate at very high neutron loads. In steady-state reactors the minimum fluence before the scheduled replacement of the reactor core should be at least l0-15 Mw.yr/m2. A more frequent replacement of the core is hardly compatible with economic constraints. A most recent summary of the discussions of these issues is presented in Ref. [l]. If and when times come to build a commercial fusion reactor, the availability of information on the behavior of materials and components at such fluences will become mandatory for making a final decision. This makes it necessary an early development and construction of a neutron source for fusion material and component testing. In this paper, we present information on one very attractive concept of such a source: a source based on a so called Gas Dynamic Trap. This neutron source was proposed in the mid 1980s (Ref. [2]; see also a survey [3] with discussion of the early stage of the project). Since then, gradual accumulation of the relevant experimental information on a modest-scale experimental facility GDT at Novosibirsk, together with a continuing design activity, have made initial theoretical considerations much more credible. We believe that such a source can be built within 4 or 5 years. Of course, one should remember that there is a chance for developing steady-state reactors with a liquid (and therefore continuously renewable) first wall [4], which would also serve as a tritium breeder. In this case, the need in the neutron testing will become less pressing. However, it is not clear yet that the concept of the flowing wall will be compatible with all types of steady-state reactors. It seems therefore prudent to be prepared to the need of a quick construction of a neutron source. It should also be mentioned that there exist projects of the accelerator-based neutron sources (e.g., [5]). However, they generally have two major disadvantages: a wrong neutron spectrum

  8. Ion engine endurance testing at high background pressures

    NASA Technical Reports Server (NTRS)

    Brophy, John R.; Pless, Lewis C.; Garner, Charles E.

    1992-01-01

    Ion engine endurance testing at vacuum chamber pressures in the low 10 exp -3 Pa range is enabled through the use of a three-grid accelerator system with the decelerator grid biased 50 to 100 volts negative of neutralizer cathode potential. The negative decelerator grid serves to collect the facility induced charge exchange ion current which normally results in rapid erosion of the accelerator grid during testing at elevated vacuum chamber pressures. This screen, accelerator, negative decelerator (SAND) grid configuration enables an order of magnitude reduction in vacuum chamber pumping speeds relative to that required for endurance testing of ion engines with conventional two-grid accelerator systems. A 900-hr test of a 30-cm diameter engine at 6.5 kW and a tank pressure of 3.7 x 10 exp -3 Pa was performed to test the feasibility of the three-grid SAND accelerator system technology. Grid erosion rates from this test are compared to those from a 200-hr test performed with the same discharge chamber, in the same test facility, and at the same background pressure with a conventional two grid accelerator system. The SAND optics resulted in greater than a factor of 100 reduction in accelerator grid erosion rate relative to the two-grid system.

  9. Ion engine endurance testing at high background pressures

    NASA Astrophysics Data System (ADS)

    Brophy, John R.; Pless, Lewis C.; Garner, Charles E.

    1992-07-01

    Ion engine endurance testing at vacuum chamber pressures in the low 10 exp -3 Pa range is enabled through the use of a three-grid accelerator system with the decelerator grid biased 50 to 100 volts negative of neutralizer cathode potential. The negative decelerator grid serves to collect the facility induced charge exchange ion current which normally results in rapid erosion of the accelerator grid during testing at elevated vacuum chamber pressures. This screen, accelerator, negative decelerator (SAND) grid configuration enables an order of magnitude reduction in vacuum chamber pumping speeds relative to that required for endurance testing of ion engines with conventional two-grid accelerator systems. A 900-hr test of a 30-cm diameter engine at 6.5 kW and a tank pressure of 3.7 x 10 exp -3 Pa was performed to test the feasibility of the three-grid SAND accelerator system technology. Grid erosion rates from this test are compared to those from a 200-hr test performed with the same discharge chamber, in the same test facility, and at the same background pressure with a conventional two grid accelerator system. The SAND optics resulted in greater than a factor of 100 reduction in accelerator grid erosion rate relative to the two-grid system.

  10. 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.

  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. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2012-12-01

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

  14. 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.

  15. 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.

  16. 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.

  17. 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

  18. 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.

  19. 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.

  20. 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

  1. STARTUP REACTIVITY ACCOUNTABILITY ATTRIBUTED TO ISOTOPIC TRANSMUTATIONS IN THE IRRADIATED BERYLLIUM REFLECTOR OF THE HIGH FLUX ISTOTOPE REACTOR

    SciTech Connect

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

    2010-01-01

    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. The computer program 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.

  2. The CG-1D neutron imaging beamline at the Oak Ridge National Laboratory High Flux Isotope Reactor

    SciTech Connect

    Santodonato, Louis J; Bilheux, Hassina Z; Bailey, William Barton; Bilheux, Jean-Christophe; Nguyen, Phong T; Tremsin, Anton S; Selby, Douglas L; Walker, Lakeisha MH

    2015-01-01

    The Oak Ridge National Laboratory Neutron Sciences Directorate has installed a neutron imaging beamline at the High Flux Isotope Reactor (HFIR) cold guide hall. CG-1D is one of the three instruments that make up the CG1 instrument suite. The beamline optics and detector have recently been upgraded to meet the needs of the neutron imaging community (better smoothing of guide system artifacts, higher flux or spatial resolution). These upgrades comprise a new diffuser/aperture system, two new detectors, a He-filled flight tube and silicon (Si) windows. Shielding inside the flight tube, beam scrapers and a beam stop ensure that biological dose is less than 50 Sv/hr outside of the radiation boundary. A set of diffusers and apertures (pinhole geometry) has been installed at the exit of the guide system to allow motorized L/D variation. Samples sit on a translation/rotation stage for alignment and tomography purposes. Detectors for the CG-1D beamline are (1) an ANDOR DW936 charge coupled device (CCD) camera with a field of view of approximately 7 cm x 7 cm and ~ 80 microns spatial resolution and 1 frame per second time resolution, (2) a new Micro-Channel Plate (MCP) detector with a 2.8 cm x 2.8 cm field of view and 55 microns spatial resolution, and 5 s timing capability. 6LiF/ZnS scintillators of thickness varying from 50 to 200 microns are being used at this facility. An overview of the beamline upgrade and preliminary data is presented here.

  3. The CG-1D Neutron Imaging Beamline at the Oak Ridge National Laboratory High Flux Isotope Reactor

    NASA Astrophysics Data System (ADS)

    Santodonato, Lou; Bilheux, Hassina; Bailey, Barton; Bilheux, Jean; Nguyen, Phong; Tremsin, Anton; Selby, Doug; Walker, Lakeisha

    The Oak Ridge National Laboratory Neutron Sciences Directorate has installed a neutron imaging beamline at the High Flux Isotope Reactor (HFIR) cold guide hall. CG-1D is one of the three instruments that make up the CG1 instrument suite. The beamline optics and detector have recently been upgraded to meet the needs of the neutron imaging community (better "smoothing" of guide system artifacts, higher flux or spatial resolution). These upgrades comprise a new diffuser/aperture system, two new detectors, a He-filled flight tube and silicon (Si) windows. Shielding inside the flight tube, beam scrapers and a beam stop ensure that biological dose is less than 50 μSv/hr outside of the radiation boundary. A set of diffusers and apertures (pinhole geometry) has been installed at the exit of the guide system to allow motorized L/D variation. Samples sit on a translation/rotation stage for alignment and tomography purposes. Detectors for the CG-1D beamline are (1) an ANDOR DW936 charge coupled device (CCD) camera with a field of view of approximately 7 cm x 7 cm and ∼ 80 microns spatial resolution and 1 frame per second time resolution, (2) a new Micro-Channel Plate (MCP) detector with a 2.8 cm x 2.8 cm field of view and 55 microns spatial resolution, and 5 μs timing capability. 6LiF/ZnS scintillators of thickness varying from 50 to 200 microns are being used at this facility. An overview of the beamline upgrade and preliminary data is presented here.

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

    SciTech Connect

    Selby, D.

    1998-05-01

    The scope of this project includes the development, design, procurement/fabrication, testing, and installation of all of the components necessary to produce a working cold source within an existing HFIR beam tube hole in the pressure vessel. All aspects of the cold source design will be based on demonstrated technology adapted to the HFIR design and operating conditions.

  5. Hydrogen loops in existing reactors for testing fuel elements for nuclear propulsion

    NASA Astrophysics Data System (ADS)

    Olsen, Charles S.; Welland, Henry; Abraschoff, James; Thoms, Kenneth

    1993-01-01

    The Space Exploration Initiative (SEI) has revitalized interest in adapting nuclear energy for power and propulsion. Prior to the selection of a nuclear thermal propulsion (NTP) system, extensive testing of the various proposed concepts will be required. In today's environmental, safety and health culture, full size rocket engine tests as were done under the Rover/NERVA program will be extremely difficult and expensive to perform and meet NASA's schedules. A different test strategy uses a hydrogen loop in an existing reactor to test a wide variety of single elements or clusters of elements for fuel qualification. This approach is expected to reduce operating and capital costs and expedite the testing schedule. This paper examines the potential of performing subscale tests in a hydrogen loop in an existing reactor 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. The HFIR is expected to achieve power densities comparable to those achieved in ATR because of the 85 MWt power level and the high thermal and fast flux levels. The available length and diameter of the test region of FHIR are 60 cm and 10 cm whereas the available length and diameter of the test region of ATR are 120 cm and 12 cm respectively.

  6. Hydrogen loops in existing reactors for testing fuel elements for nuclear propulsion

    SciTech Connect

    Olsen, C.S.; Welland, H.; Abraschoff, J. ); Thoms, K. )

    1993-01-15

    The Space Exploration Initiative (SEI) has revitalized interest in adapting nuclear energy for power and propulsion. Prior to the selection of a nuclear thermal propulsion (NTP) system, extensive testing of the various proposed concepts will be required. In today's environmental, safety and health culture, full size rocket engine tests as were done under the Rover/NERVA program will be extremely difficult and expensive to perform and meet NASA's schedules. A different test strategy uses a hydrogen loop in an existing reactor to test a wide variety of single elements or clusters of elements for fuel qualification. This approach is expected to reduce operating and capital costs and expedite the testing schedule. This paper examines the potential of performing subscale tests in a hydrogen loop in an existing reactor 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. The HFIR is expected to achieve power densities comparable to those achieved in ATR because of the 85 MWt power level and the high thermal and fast flux levels. The available length and diameter of the test region of FHIR are 60 cm and 10 cm whereas the available length and diameter of the test region of ATR are 120 cm and 12 cm respectively.

  7. 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.

  8. Testing of Gas Reactor Fuel and Materials in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2006-10-01

    The recent growth in interest for high temperature gas reactors has resulted in an increased need for materials and fuel testing for this type of reactor. The Advanced Test Reactor (ATR), located at the US Department of Energy’s Idaho National Laboratory, has long been involved in testing gas reactor fuel and materials, and has facilities and capabilities to provide the right environment for gas reactor irradiation experiments. These capabilities include both passive sealed capsule experiments, and instrumented/actively controlled experiments. The instrumented/actively controlled experiments typically contain thermocouples and control the irradiation temperature, but on-line measurements and controls for pressure and gas environment have also been performed in past irradiations. The ATR has an existing automated gas temperature control system that can maintain temperature in an irradiation experiment within very tight bounds, and has developed an on-line fission product monitoring system that is especially well suited for testing gas reactor particle fuel. The ATR’s control system, which consists primarily of vertical cylinders used to rotate neutron poisons/reflectors toward or away from the reactor core, provides a constant vertical flux profile over the duration of each operating cycle. This constant chopped cosine shaped axial flux profile, with a relatively flat peak at the vertical centre of the core, is more desirable for experiments than a constantly moving axial flux peak resulting from a control system of axially positioned control components which are vertically withdrawn from the core.

  9. Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

    SciTech Connect

    D. L. Knudson; J. L. Rempe

    2012-02-01

    New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine, and return irradiated samples for each measurement makes this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated in pressurized water reactor (PWR) coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL's High Temperature Test Laboratory (HTTL).

  10. Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

    SciTech Connect

    D. L. Knudson; J. L. Rempe

    2012-02-01

    New materials are being considered for fuel, cladding and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine and return irradiated samples for each measurement make this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated under pressurized water reactor coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory.

  11. 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.

  12. 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.

  13. High heat flux testing of CFC composites for the tokamak physics experiment

    NASA Astrophysics Data System (ADS)

    Valentine, P. G.; Nygren, R. E.; Burns, R. W.; Rocket, P. D.; Colleraine, A. P.; Lederich, R. J.; Bradley, J. T.

    1996-10-01

    High heat flux (HHF) testing of carbon fiber reinforced carbon composites (CFC's) was conducted under the General Atomics program to develop plasma-facing components (PFC's) for Princeton Plasma Physics Laboratory's tokamak physics experiment (TPX). As part of the process of selecting TPX CFC materials, a series of HHF tests were conducted with the 30 kW electron beam test system (EBTS) facility at Sandia National Laboratories, and with the plasma disruption simulator I (PLADIS-I) facility at the University of New Mexico. The purpose of the tests was to make assessments of the thermal performance and erosion behavior of CFC materials. Tests were conducted with 42 different CFC materials. In general, the CFC materials withstood the rapid thermal pulse environments without fracturing, delaminating, or degrading in a non-uniform manner; significant differences in thermal performance, erosion behavior, vapor evolution, etc. were observed and preliminary findings are presented below. The CFC's exposed to the hydrogen plasma pulses in PLADIS-I exhibited greater erosion rates than the CFC materials exposed to the electron-beam pulses in EBTS. The results obtained support the continued consideration of a variety of CFC composites for TPX PFC components.

  14. Large break loss of coolant severe accident sequences at the HFIR (High Flux Isotope Reactor)

    SciTech Connect

    Simpson, D.B.; Greene, S.R.

    1990-01-01

    An assessment of many potential HFIR severe accident phenomena was conducted during the HFIR design effort, and many severe accident mitigating features were designed into the plant. These evaluation typically incorporated a bounding'' or highly conservative analysis approach and employed tools and techniques representative of the state of knowledge in the mid-1960s. Recently, programs to address severe accident issues were initiated at the Oak Ridge National Laboratory (ORNL) to support the HFIR probabilistic risk assessment (PRA) and equipment qualification and accident management studies. This paper presents the results of environment condition calculations conducted to evaluate a response of HFIR's heat exchanger cell environment to a double-ended rupture of a 0.25 m diameter coolant loop downstream of the circulating pump and check valve. The confinement calculations were performed using an atmospheric fission product source for the heat exchanger cell consistent with, but more conservative than that stipulated in Regulatory Guide 1.89. The results of the calculations indicate that the heat exchanger cell atmospheric temperature peaks at 377 K 225 seconds into the transient and then begins decreasing at approximately 1.7 K per minute. 8 refs., 5 figs.

  15. 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.

  16. 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.

  17. 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.

  18. Advanced Test Reactor Tour

    ScienceCinema

    Miley, Don

    2013-05-28

    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.

  19. Advanced Test Reactor Tour

    SciTech Connect

    Miley, Don

    2011-01-01

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

  20. 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.

  1. Particle Bed Reactor engine technology

    SciTech Connect

    Sandler, S.; Feddersen, R. )

    1992-01-01

    This paper discusses the Particle Bed Reactor (PBR) based propulsion system being developed under the Space Nuclear Thermal Propulsion (SNTP) program. A PBR engine is a light weight, compact propulsion system which offers significant improvement over current technology systems. Current performance goals are a system thrust of 75,000 pounds at an Isp of 1000 sec. A target thrust to weight ratio (T/W) of 30 has been established for an unshielded engine. The functionality of the PBR, its pertinent technology issues and the systems required to make up a propulsion system are described herein. Accomplishments to date which include hardware development and tests for the PBR engine are also discussed. This paper is intended to provide information on and describe the current state-of-the-art of PBR technology. 4 refs.

  2. 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

  3. Measurement of the Neutron Spectrum of the HB-4 Cold Source at the High Flux Isotope Reactor at Oak Ridge National Laboratory

    NASA Astrophysics Data System (ADS)

    Robertson, J. L.; Iverson, E. B.

    2009-08-01

    Measurements of the cold neutron spectrum from the super critical hydrogen cold source at the High Flux Isotope Reactor at Oak Ridge National Laboratory were made using time-of-flight spectroscopy. Data were collected at reactor power levels of 8.5MW, 42.5MW and 85MW. The moderator temperature was also varied. Data were collected at 17K and 25K while the reactor power was at 8.5MW, 17K and 25K while at 42.5MW and 18K and 22K while at 85MW. The purpose of these measurements was to characterize the brightness of the cold source and to better understand the relationship between reactor power, moderator temperature, and cold neutron production. The authors will discuss the details of the measurement, the changes observed in the neutron spectrum, and the process for determining the source brightness from the measured neutron intensity.

  4. 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.

  5. 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).

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. Development of micro-engineered textured tungsten surfaces for high heat flux applications

    NASA Astrophysics Data System (ADS)

    Sharafat, Shahram; Aoyama, Aaron; Williams, Brian; Ghoniem, Nasr

    2013-11-01

    Surface micro-engineering can enhance the thermo-mechanical performance of plasma facing components (PFCs). For example, castellation of a surface can reduce thermal stress due to high heat loads and thus provide higher thermo-mechanical resilience. Recently, fabrication of a variety of micro-sized refractory dendrites with reproducible geometric characteristics (e.g., density, length, height, and aspect ratio) has been demonstrated. In contrast to flat surfaces exposed to high heat loads, dendrites deform independently to minimize near-surface thermal stress, which results in improved thermo-mechanical performance. Thus, the use of dendrites offers a unique micro-engineering approach to enhance the performance of PFC structures. A brief overview of W, Re, and Mo dendritic structures is given along with micrographs that show dendrite-coated surfaces. The thermal responses of representative dendrite structures are analyzed as a function of aspect ratios and dendrite geometry. The heat-management capability of needle-like dendrites exposed to a surface energy of up to 1 MJ/m2 is analyzed and compared to a flat surface. It is concluded that dendrite structures can significantly reduce thermal stress in the substrate when compared to flat surfaces. Implications of dendritic surfaces on sputter erosion rates are also discussed briefly. Higher heat load capability, due to increased surface area and quasi-volumetric heat deposition instead of purely surface loading. Reduced potential for crack initiation, due to the presence of the dendrites, which absorb most of the heat and then conduct it to the substrate. In effect, the dendritic surface is the ultimate in "castellated" (tiled) surfaces, which reduce thermal stresses and thus cracking. Reduced subsurface implantation depth of high-energy helium. A high aspect ratio (height/diameter) dendrite has very shallow ion impingement angles along the sides, which results in a significant reduction in the depth of penetration of

  12. 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.

  13. 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.

  14. Cross section generation and physics modeling in a feasibility study of the conversion of the high flux isotope reactor core to use low-enriched uranium fuel

    SciTech Connect

    Ellis, R. J.; Gehin, J. C.; Primm Iii, R. T.

    2006-07-01

    A computational study has been initiated at ORNL to examine the feasibility of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The current study is limited to steady-state, nominal operation and are focused on the determination of the fuel requirements, primarily density, that are required to maintain the performance of the reactor. Reactor physics analyses are reported for a uranium-molybdenum alloy that would be substituted for the current fuel - U{sub 3}O{sub 8} mixed with aluminum. An LEU core design has been obtained and requires an increase in {sup 235}U loading of a factor of 1.9 over the current HEU fuel. These initial results indicate that the conversion from HEU to LEU results in a reduction of the thermal fluxes in the central flux trap region of approximately 9 % and in the outer beryllium reflector region of approximately 15%. Ongoing work is being performed to improve upon this initial design to further minimize the impact of conversion to LEU fuel. (authors)

  15. 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.

  16. 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.

  17. Development and Testing of a High Stability Engine Control (HISTEC) System

    NASA Technical Reports Server (NTRS)

    Orme, John S.; DeLaat, John C.; Southwick, Robert D.; Gallops, George W.; Doane, Paul M.

    1998-01-01

    Flight tests were recently completed to demonstrate an inlet-distortion-tolerant engine control system. These flight tests were part of NASA's High Stability Engine Control (HISTEC) program. The objective of the HISTEC program was to design, develop, and flight demonstrate an advanced integrated engine control system that uses measurement-based, real-time estimates of inlet airflow distortion to enhance engine stability. With improved stability and tolerance of inlet airflow distortion, future engine designs may benefit from a reduction in design stall-margin requirements and enhanced reliability, with a corresponding increase in performance and decrease in fuel consumption. This paper describes the HISTEC methodology, presents an aircraft test bed description (including HISTEC-specific modifications) and verification and validation ground tests. Additionally, flight test safety considerations, test plan and technique design and approach, and flight operations are addressed. Some illustrative results are presented to demonstrate the type of analysis and results produced from the flight test program.

  18. Test Pilot and Engineer

    NASA Technical Reports Server (NTRS)

    1920-01-01

    Goggles at the ready, this Langley test pilot and engineer conducted research business high above the ground. In the early years the flight research team was usually made up of a test pilot (Thomas Carroll, front cockpit) and an engineer (John W. Gus Crowley, Jr.).

  19. Facility for high heat flux testing of irradiated fusion materials and components using infrared plasma arc lamps

    SciTech Connect

    Sabau, Adrian S; Ohriner, Evan Keith; Kiggans, Jim; Harper, David C; Snead, Lance Lewis; Schaich, Charles Ross

    2014-01-01

    A new high-heat flux testing facility using water-wall stabilized high-power high-pressure argon Plasma Arc Lamps (PALs) has been developed for fusion applications. It can handle irradiated plasma facing component materials and mock-up divertor components. Two PALs currently available at ORNL can provide maximum incident heat fluxes of 4.2 and 27 MW/m2 over a heated area of 9x12 and 1x10 cm2, respectively, which are fusion-prototypical steady state heat flux conditions. The facility will be described and the main differences between the photon-based high-heat flux testing facilities, such as PALs, and the e-beam and particle beam facilities more commonly used for fusion HHF testing are discussed. The components of the test chamber were designed to accommodate radiation safety and materials compatibility requirements posed by high-temperature exposure of low levels irradiated tungsten articles. Issues related to the operation and temperature measurements during testing are presented and discussed.

  20. Production of medical radioisotopes in the ORNL high flux isotope reactor (HFIR) for cancer treatment and arterial restenosis therapy after PICA

    NASA Astrophysics Data System (ADS)

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

    1999-01-01

    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. First beginning operation in 1965, the high thermal neutron flux (2.5×1015 neutrons/cm2/sec at 85 MW) and versatile target irradiation and handling facilities provide the opportunity for production of a wide variety of neutron-rich medical radioisotopes of current interest for therapy. 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-117 m 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 (22-24 days) 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-117 m 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.

  1. Production of medical radioisotopes in the ORNL high flux isotope reactor (HFIR) for cancer treatment and arterial restenosis therapy after PICA

    NASA Astrophysics Data System (ADS)

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

    1999-01-01

    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. First beginning operation in 1965, the high thermal neutron flux (2.5×1015 neutrons/cm2/sec at 85 MW) and versatile target irradiation and handling facilities provide the opportunity for production of a wide variety of neutron-rich medical radioisotopes of current interest for therapy. 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 (22 24 days) 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.

  2. 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…

  3. Startup Testing of the Fast Flux Test Facility

    SciTech Connect

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

    2010-06-30

    This paper is one in a series documenting the current effort to retrieve, secure, and preserve critical information related to advanced reactors. . Information from this testing is being retrieved under the Fuel Cycle Research and Development (FCRD) program conducted by the Office of Nuclear Energy (NE) of the DOE. The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE).

  4. Design of a high-flux epithermal neutron beam using 235U fission plates at the Brookhaven Medical Research Reactor.

    PubMed

    Liu, H B; Brugger, R M; Rorer, D C; Tichler, P R; Hu, J P

    1994-10-01

    Beams of epithermal neutrons are being used in the development of boron neutron capture therapy for cancer. This report describes a design study in which 235U fission plates and moderators are used to produce an epithermal neutron beam with higher intensity and better quality than the beam currently in use at the Brookhaven Medical Research Reactor (BMRR). Monte Carlo calculations are used to predict the neutron and gamma fluxes and absorbed doses produced by the proposed design. Neutron flux measurements at the present epithermal treatment facility (ETF) were made to verify and compare with the computed results where feasible. The calculations indicate that an epithermal neutron beam produced by a fission-plate converter could have an epithermal neutron intensity of 1.2 x 10(10) n/cm2.s and a fast neutron dose per epithermal neutron of 2.8 x 10(-11) cGy.cm2/nepi plus being forward directed. This beam would be built into the beam shutter of the ETF at the BMRR. The feasibility of remodeling the facility is discussed. PMID:7869995

  5. 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.

  6. Advanced Test Reactor Testing Experience: Past, Present and Future

    SciTech Connect

    Frances M. Marshall

    2005-04-01

    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 48" long and 5.0" diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors -- US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, wherein the target material is placed in a capsule, or plate form, and the capsule is in direct contact with the primary coolant. The next level of complexity of an experiment is an instrumented lead experiment, which allows for active monitoring and control of experiment conditions during the irradiation. The highest level of complexity of experiment is the pressurized water loop experiment, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans.

  7. Reactor antineutrino fluxes - Status and challenges

    NASA Astrophysics Data System (ADS)

    Huber, Patrick

    2016-07-01

    In this contribution we describe the current understanding of reactor antineutrino fluxes and point out some recent developments. This is not intended to be a complete review of this vast topic but merely a selection of observations and remarks, which despite their incompleteness, will highlight the status and the challenges of this field.

  8. 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

  9. Helium-cooled divertor for DEMO: Manufacture and high heat flux tests of tungsten-based mock-ups

    NASA Astrophysics Data System (ADS)

    Norajitra, P.; Gervash, A.; Giniyatulin, R.; Hirai, T.; Janeschitz, G.; Krauss, W.; Kuznetsov, V.; Makhankov, A.; Mazul, I.; Ovchinnikov, I.; Reiser, J.; Widak, V.

    2009-04-01

    A helium-cooled divertor concept for DEMO has been investigated extensively at the Forschungszentrum Karlsruhe under the EU power plant conceptual study, the goal being to demonstrate performance under heat flux of 10 MW/m 2 at least. Work covers different areas ranging from conceptual design to analysis, materials and fabrication issues to experiments. Meanwhile, the He-cooled modular divertor concept with jet cooling (HEMJ) has been proposed as reference design. In cooperation with the Efremov Institute, manufacture and high heat flux testing of divertor elements was performed for design verification and proof-of-principle. This paper focuses on the technological study of the fabrication of mock-ups from W/W alloy and Eurofer steel supporting structure material. The high heat flux test results of 2006 and 2007 are summarised and discussed.

  10. 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.

  11. Efficient testing of ITER materials and components at the Research Institute of Atomic Reactors` experimental facilities

    SciTech Connect

    Ivanov, V.; Kazakov, V.; Pokrovsky, A.; Shamardin, V.; Melder, R.; Revyakin, Yu.; Sandakov, V.

    1995-12-31

    The Research Institute of Atomic Reactors (RIAR) of the State Scientific Centre of the Russian Federation has carried out reactor tests of fusion reactor materials and components. RIAR contains an ideal complex of installations, experimental setups, and diagnostics for such investigations. It includes several different types of reactors, including a fast neutron reactor, a high-flux intermediate-neutron SM-3 reactor, a intermediate-neutron loop reactor, and two RBT-type reactors, and a hot cells complex with remote handling facilities to allow study of the physical-mechanical properties, structure, and elemental composition of irradiated materials. RIAR has carried out a number of initial experiments, including testing of copper and vanadium alloys, electro-insulative coatings, steels, ceramics, diagnostic systems materials, and in-core and hot cell set-ups for divertor mock-up testing, and has collaborative efforts underway with the Scientific Research Institute Electrophysical Apparatus-St. Petersburg (SRIEA), Oak Ridge National Laboratory (ORNL), Argonne National Laboratory (ANL), Red Star, the Institute of Physics and Power engineering (IPPE), the Scientific Research Institute of Inorganic Materials (SRIIM), and Pacific Northwest Laboratory (PNL).

  12. Facility for high-heat flux testing of irradiated fusion materials and components using infrared plasma arc lamps

    NASA Astrophysics Data System (ADS)

    Sabau, Adrian S.; Ohriner, Evan K.; Kiggans, Jim; Harper, David C.; Snead, Lance L.; Schaich, Charles R.

    2014-04-01

    A new high-heat flux testing (HHFT) facility using water-wall stabilized high-power high-pressure argon plasma arc lamps (PALs) has been developed for fusion applications. It can accommodate irradiated plasma facing component materials and sub-size mock-up divertor components. Two PALs currently available at Oak Ridge National Laboratory can provide maximum incident heat fluxes of 4.2 and 27 MW m-2, which are prototypic of fusion steady state heat flux conditions, over a heated area of 9 × 12 and 1 × 10 cm2, respectively. The use of PAL permits the heat source to be environmentally separated from the components of the test chamber, simplifying the design to accommodate safe testing of low-level irradiated articles and materials under high-heat flux. Issues related to the operation and temperature measurements during testing of tungsten samples are presented and discussed. The relative advantages and disadvantages of this photon-based HHFT facility are compared to existing e-beam and particle beam facilities used for similar purposes.

  13. SUMMARY AND RESULTS LETTER REPORT – INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PROJECT, PHASE 3: TRENCHES 2, 3, AND 4 BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    SciTech Connect

    E.M. Harpenau

    2010-11-15

    5098-LR-02-0 SUMMARY AND RESULTS LETTER REPORT – INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PROJECT, PHASE 3 TRENCHES 2, 3, AND 4 BROOKHAVEN NATIONAL LABORATORY

  14. Metal fuel test program in the Fast Flux Test Facility

    SciTech Connect

    Pitner, A.L.; Baker, R.B.

    1992-10-01

    This report discusses irradiation testing of metal fuel assemblies in the Fast Flux Test Facility (FFTF) which has demonstrated the viability of this robust fuel design for liquid metal reactor applications. This fuel design provides high burnup capability with reduced fabrication costs relative to standard mixed-oxide FFTF driver fuel assemblies. Development of this fuel design required the establishment of innovative sodium bonding technology as well as special techniques for sodium bond quality verification. Eight metal fuel test assemblies have been irradiated under demanding conditions to burnups as high as 143 MWd/kgM with no indication of pin breach. The unique FFTF instrumentation system has permitted the in situ observation of axial fuel growth in metal fuel assemblies.

  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. 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.

  17. Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

    SciTech Connect

    T. A. Tomberlin; S. B. Grover

    2004-11-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

  18. 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.

  19. 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.

  20. Thermal-hydraulic simulation of natural convection decay heat removal in the High Flux Isotope Reactor using RELAP5 and TEMPEST: Part 1, Models and simulation results

    SciTech Connect

    Morris, D.G.; Wendel, M.W.; Chen, N.C.J.; Ruggles, A.E.; Cook, D.H.

    1989-01-01

    A study was conducted to examine decay heat removal requirements in the High Flux Isotope Reactor (HFIR) following shutdown from 85 MW. The objective of the study was to determine when forced flow through the core could be terminated without causing the fuel to melt. This question is particularly relevant when a station blackout caused by an external event is considered. Analysis of natural circulation in the core, vessel upper plenum, and reactor pool indicates that 12 h of forced flow will permit a safe shutdown with some margin. However, uncertainties in the analysis preclude conclusive proof that 12 h is sufficient. As a result of the study, two seismically qualified diesel generators were installed in HFIR. 9 refs., 4 figs.

  1. 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.

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

    DOE PAGESBeta

    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 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

  3. 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.

  4. 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 been 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.

  5. 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.

  6. Mechanical properties and microstructure of copper alloys and copper alloy-stainless steel laminates for fusion reactor high heat flux applications

    NASA Astrophysics Data System (ADS)

    Leedy, Kevin Daniel

    A select group of copper alloys and bonded copper alloy-stainless steel panels are under consideration for heat sink applications in first wall and divertor structures of a planned thermonuclear fusion reactor. Because these materials must retain high strengths and withstand high heat fluxes, their material properties and microstructures must be well understood. Candidate copper alloys include precipitate strengthened CuNiBe and CuCrZr and dispersion strengthened Cu-Alsb2Osb3 (CuAl25). In this study, uniaxial mechanical fatigue tests were conducted on bulk copper alloy materials at temperatures up to 500sp°C in air and vacuum environments. Based on standardized mechanical properties measurement techniques, a series of tests were also implemented to characterize copper alloy-316L stainless steel joints produced by hot isostatic pressing or by explosive bonding. The correlation between mechanical properties and the microstructure of fatigued copper alloys and the interface of copper alloy-stainless steel laminates was examined. Commercial grades of these alloys were used to maintain a degree of standardization in the materials testing. The commercial alloys used were OMG Americas Glidcop CuAl25 and CuAl15; Brush Wellman Hycon 3HP and Trefimetaux CuNiBe; and Kabelmetal Elbrodur and Trefimetaux CuCrZr. CuAl25 and CuNiBe alloys possessed the best combination of fatigue resistance and microstructural stability. The CuAl25 alloy showed only minimal microstructural changes following fatigue while the CuNiBe alloy consistently exhibited the highest fatigue strength. Transmission electron microscopy observations revealed that small matrix grain sizes and high densities of submicron strengthening phases promoted homogeneous slip deformation in the copper alloys. Thus, highly organized fatigue dislocation structure formation, as commonly found in oxygen-free high conductivity Cu, was inhibited. A solid plate of CuAl25 alloy hot isostatically pressed to a 316L stainless steel

  7. 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

  8. 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.

  9. 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.

  10. Test pilot and engineer

    NASA Technical Reports Server (NTRS)

    1922-01-01

    Goggles at the ready, this Langley test pilot and engineer conducted research business high above the ground. Photograph published in Winds of Change, 75th Anniversary NASA publication, by James Schultz (page 24). This photograph is also published in Engineer in Charge: A History of the Langley Aeronautical Laboratory, 1917-1958 by James R. Hansen (page 163). In the early years the flight research team was usually made up of a test pilot (Thomas Carroll, front cockpit) and an engineer (John W. Gus Crowley,Jr.).

  11. 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

  12. Nuclear Reactor Engineering Analysis Laboratory

    SciTech Connect

    Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

    1998-12-31

    The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels.

  13. 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.

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

    DOE PAGESBeta

    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 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

  15. 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.

  16. Optimization of a partially non-magnetic primary radiation. shielding for the triple-axis spectrometer PANDA at the Munich high-flux reactor FRM-II

    NASA Astrophysics Data System (ADS)

    Pyka, N. M.; Noack, K.; Rogov, A.

    Monte Carlo simulations have been used to optimize the monochromator shielding of the polarized cold-neutron triple-axis spectrometer PANDA at the Munich high-flux reactor FRM-II. By using the Monte Carlo program MCNP-4B, the density of the total spectrum of incoming neutrons and γ radiation from the beam tube SR-2 has been determined during the three-dimensional diffusion process in different types of heavy concrete and other absorbing material. Special attention has been paid to build a compact and highly efficient shielding, partially non-magnetic, with a total biological radiation dose of less than 10 μSv/h at its outsides. Especially considered was the construction of an albedo reducer, which serves to reduce the background in the experiment outside the shielding.

  17. 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.

  18. 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.

  19. Department of Energy's High Flux Beam Reactor (HFBR), September 15--19, 1980: An independent on-site safety review

    SciTech Connect

    Not Available

    1981-02-01

    The intent of this on-site safety review was to make a broad management assessment of HFBR operations, rather than conduct a detailed in-depth audit. The result of the review should only be considered as having identified trends or indications. The Team's observations and recommendations for the most part are based upon licensed reactor facility practices used to meet industry standards. These standards form the basis for many of the comments in this report. The Team believes that a uniform minimum standard of performance should be achieved in the operation of DOE reactors. In order to assure that this is accomplished, clear standards are necessary. Consistent with the past AEC and ERDA policy, the team has used the standards of the commercial nuclear power industry. It is recognized that this approach is conservative in that the HFBR reactor has a significantly greater degree of inherent safety (low pressure, temperature, power, etc.) than a licensed reactor.

  20. 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.

  1. Transfer Calibration Validation Tests on a Heat Flux Sensor in the 51 mm High-Temperature Blackbody

    PubMed Central

    Murthy, A. V.; Tsai, B. K.; Saunders, R. D.

    2001-01-01

    Facilities and techniques to characterize heat flux sensors are under development at the National Institute of Standards and Technology. As a part of this effort, a large aperture high-temperature blackbody was commissioned recently. The graphite tube blackbody, heated electrically, has a cavity diameter of 51 mm and can operate up to a maximum temperature of 2773 K. A closed-loop cooling system using a water-to-water heat exchanger cools electrodes and the outer reflecting shield. This paper describes the newly developed blackbody facility and the validation tests conducted using a reference standard Schmidt-Boelter heat flux sensor. The transfer calibration results obtained on the Schmidt-Boelter sensor agreed with the previous data within the experimental uncertainty limits.

  2. 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.

  3. Knowledge Preservation at the Fast Flux Test Facility

    SciTech Connect

    Wootan, David W.; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.; Nelson, Joseph V.; Polzin, David L.

    2011-11-30

    The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to operate in the United States, from 1982 to 1992. The technologies employed in designing and constructing this reactor, along with information obtained from tests conducted during its operation, are currently being secured and archived by the Department of Energy's Office of Nuclear Energy. This report provides a status update documenting the overall project efforts to retrieve and preserve critical information related to advanced reactors.

  4. Knowledge Preservation at the Fast Flux Test Facility

    SciTech Connect

    Wootan, David W.; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.; Nelson, Joseph V.; Polzin, David L.

    2012-01-30

    The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to operate in the United States, from 1982 to 1992. The technologies employed in designing and constructing this reactor, along with information obtained from tests conducted during its operation, are currently being secured and archived by the Department of Energy's Office of Nuclear Energy. This report is one in a series documenting the overall project efforts to retrieve and preserve critical information related to advanced reactors

  5. 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).

  6. 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.

  7. 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)

  8. 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.

  9. 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.

  10. 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

  11. 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

  12. 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

  13. Russian Rocket Engine Test

    NASA Technical Reports Server (NTRS)

    1998-01-01

    NASA engineers successfully tested a Russian-built rocket engine on November 4, 1998 at the Marshall Space Flight Center (MSFC) Advanced Engine Test Facility, which had been used for testing the Saturn V F-1 engines and Space Shuttle Main engines. The MSFC was under a Space Act Agreement with Lockheed Martin Astronautics of Denver to provide a series of test firings of the Atlas III propulsion system configured with the Russian-designed RD-180 engine. The tests were designed to measure the performance of the Atlas III propulsion system, which included avionics and propellant tanks and lines, and how these components interacted with the RD-180 engine. The RD-180 is powered by kerosene and liquid oxygen, the same fuel mix used in Saturn rockets. The RD-180, the most powerful rocket engine tested at the MSFC since Saturn rocket tests in the 1960s, generated 860,000 pounds of thrust.

  14. 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.

  15. Porous Structure Analysis of the Packed Beds in a High-Temperature Reactor Pebble Bed Modules Heat Transfer Test Facility

    NASA Astrophysics Data System (ADS)

    Ren, Cheng; Yang, Xing-Tuan; Sun, Yan-Fei

    2013-02-01

    We analyse the porous structure of the packed beds in the heat transfer test facility built for high temperature gas cooled reactors from several aspects, such as oscillatory porosity, average porosity, thickness effect, coordination number and contact angle. An understanding and comparison of the porous structure of the facility bed and the real reactor core are developed to make recommendations for the design and analysis of the heat transfer test facility. The results show that there is very little difference between the porous characteristics of the two packed beds of spheres.

  16. Fabrication, Inspection, and Test Plan for the Advanced Test Reactor (ATR) High-Power Mixed-Oxide (MOX) Fuel Irradiation Project

    SciTech Connect

    Wachs, G. W.

    1998-09-01

    The Department of Energy (DOE) Fissile Disposition Program (FMDP) has announced that reactor irradiation of Mixed-Oxide (MOX) fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The High-Power MOX fuel test will be irradiated in the Advanced Test Reactor (ATR) to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. The purpose of the high-power experiment, in conjunction with the currently ongoing average-power experiment at the ATR, is to contribute new information concerning the response of WG plutonium under more severe irradiation conditions typical of the peak power locations in commercial reactors. In addition, the high-power test will contribute experience with irradiation of gallium-containing fuel to the database required for resolution of generic CLWR fuel design issues. The distinction between "high-power" and "average-power" relates to the position within the nominal CLWR core. The high-power test project is subject to a number of requirements, as discussed in the Fissile Materials Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation High-Power Test Project Plan (ORNL/MD/LTR-125).

  17. 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.

  18. 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.

  19. 10 CFR 707.7 - Random drug testing requirements and identification of testing designated positions.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... (HRP), codified at 10 CFR part 712. HRP employees will be subject to the drug testing standards of this... Facility (FFTF); High Flux Beam Reactor (HFBR); High Flux Isotope Reactor (HFIR); K Production Reactor...

  20. 10 CFR 707.7 - Random drug testing requirements and identification of testing designated positions.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... (HRP), codified at 10 CFR part 712. HRP employees will be subject to the drug testing standards of this... Facility (FFTF); High Flux Beam Reactor (HFBR); High Flux Isotope Reactor (HFIR); K Production Reactor...

  1. 10 CFR 707.7 - Random drug testing requirements and identification of testing designated positions.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... (HRP), codified at 10 CFR part 712. HRP employees will be subject to the drug testing standards of this... Facility (FFTF); High Flux Beam Reactor (HFBR); High Flux Isotope Reactor (HFIR); K Production Reactor...

  2. 10 CFR 707.7 - Random drug testing requirements and identification of testing designated positions.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... (HRP), codified at 10 CFR part 712. HRP employees will be subject to the drug testing standards of this... Facility (FFTF); High Flux Beam Reactor (HFBR); High Flux Isotope Reactor (HFIR); K Production Reactor...

  3. 10 CFR 707.7 - Random drug testing requirements and identification of testing designated positions.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... (HRP), codified at 10 CFR part 712. HRP employees will be subject to the drug testing standards of this... Facility (FFTF); High Flux Beam Reactor (HFBR); High Flux Isotope Reactor (HFIR); K Production Reactor...

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. Development Program of IS Process Pilot Test Plant for Hydrogen Production With High-Temperature Gas-Cooled Reactor

    SciTech Connect

    Jin Iwatsuki; Atsuhiko Terada; Hiroyuki Noguchi; Yoshiyuki Imai; Masanori Ijichi; Akihiro Kanagawa; Hiroyuki Ota; Shinji Kubo; Kaoru Onuki; Ryutaro Hino

    2006-07-01

    At the present time, we are alarmed by depletion of fossil energy and effects on global environment such as acid rain and global warming, because our lives depend still heavily on fossil energy. So, it is universally recognized that hydrogen is one of the best energy media and its demand will be increased greatly in the near future. In Japan, the Basic Plan for Energy Supply and Demand based on the Basic Law on Energy Policy Making was decided upon by the Cabinet on 6 October, 2003. In the plan, efforts for hydrogen energy utilization were expressed as follows; hydrogen is a clean energy carrier without carbon dioxide (CO{sub 2}) emission, and commercialization of hydrogen production system using nuclear, solar and biomass, not fossil fuels, is desired. However, it is necessary to develop suitable technology to produce hydrogen without CO{sub 2} emission from a view point of global environmental protection, since little hydrogen exists naturally. Hydrogen production from water using nuclear energy, especially the high-temperature gas-cooled reactor (HTGR), is one of the most attractive solutions for the environmental issue, because HTGR hydrogen production by water splitting methods such as a thermochemical iodine-sulfur (IS) process has a high possibility to produce hydrogen effectively and economically. The Japan Atomic Energy Agency (JAEA) has been conducting the HTTR (High-Temperature Engineering Test Reactor) project from the view to establishing technology base on HTGR and also on the IS process. In the IS process, raw material, water, is to be reacted with iodine (I{sub 2}) and sulfur dioxide (SO{sub 2}) to produce hydrogen iodide (HI) and sulfuric acid (H{sub 2}SO{sub 4}), the so-called Bunsen reaction, which are then decomposed endo-thermically to produce hydrogen (H{sub 2}) and oxygen (O{sub 2}), respectively. Iodine and sulfur dioxide produced in the decomposition reactions can be used again as the reactants in the Bunsen reaction. In JAEA, continuous

  9. 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.

  10. 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.

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. Neutron Flux Characterization of Irradiation Holes for Irradiation Test at HANARO

    NASA Astrophysics Data System (ADS)

    Yang, Seong Woo; Cho, Man Soon; Choo, Kee Nam; Park, Sang Jun

    2016-02-01

    The High flux Advanced Neutron Application ReactOr (HANARO) is a unique research reactor in the Republic of Korea, and has been used for irradiation testing since 1998. To conduct irradiation tests for nuclear materials, the irradiation holes of CT and OR5 have been used due to a high fast-neutron flux. Because the neutron flux must be accurately calculated to evaluate the neutron fluence of irradiated material, it was conducted using MCNP. The neutron flux was measured using fluence monitor wires to verify the calculated result. Some evaluations have been conducted, however, more than 20% errors have frequently occurred at the OR irradiation hole, while a good agreement between the calculated and measured data was shown at the CT irradiation hole.

  17. 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.

  18. High heat-flux self-rotating plasma-facing component: Concept and loading test in TEXTOR

    NASA Astrophysics Data System (ADS)

    Terra, A.; Sergienko, G.; Hubeny, M.; Huber, A.; Mertens, Ph.; Philipps, V.

    2015-08-01

    This contribution reports on the concept of a circular self-rotating and temperature self-stabilising plasma-facing component (PFC), and test of a related prototype in TEXTOR tokamak. This PFC uses the Lorentz force induced by plasma current and magnet field (J × B) to create a torque applied on metallic discs which produce a rotational movement. Additional thermionic current, present at high operation temperatures, brings additional temperature stabilisation ability. This self-rotating disk limiter was exposed to plasma in the TEXTOR tokamak under different radial positions to vary the heat flux. This disk structure shows the interesting ability to stabilise its maximum temperature through the fact that the self-induced rotation is modulated by the thermal emission current. It was observed that the rotation speed increased following both the current collected by the limiter, and the temperature of the tungsten disks.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Russian Rocket Engine Test

    NASA Technical Reports Server (NTRS)

    1998-01-01

    NASA engineers successfully tested a Russian-built rocket engine on November 4, 1998 at the Marshall Space Flight Center (MSFC) Advanced Engine Test Facility, which had been used for testing the Saturn V F-1 engines and Space Shuttle Main engines. The MSFC was under a Space Act Agreement with Lockheed Martin Astronautics of Denver to provide a series of test firings of the Atlas III propulsion system configured with the Russian-designed RD-180 engine. The tests were designed to measure the performance of the Atlas III propulsion system, which included avionics and propellant tanks and lines, and how these components interacted with the RD-180 engine. The RD-180 is powered by kerosene and liquid oxygen, the same fuel mix used in Saturn rockets. The RD-180, the most powerful rocket engine tested at the MSFC since Saturn rocket tests in the 1960s, generated 860,000 pounds of thrust. The test was the first test ever anywhere outside Russia of a Russian designed and built engine.

  5. 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

  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. Preliminary Results of the Combined Third and Fourth Very High Temperature Gas-Cooled Reactor Irradiation in the Advanced Test Reactor

    SciTech Connect

    Davenport, Michael E.; Palmer, A. Joseph; Petti, David A.

    2001-10-01

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

  8. 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.

  9. Detailed flux calculations for the conceptual design of the Advanced Neutron Source Reactor

    SciTech Connect

    Wemple, C.A.

    1995-05-01

    A detailed MCNP model of the Advanced Neutron Source Reactor has been developed. All reactor components inside the reflector tank were included, and all components were highly segmented. Neutron and photon multigroup flux spectra have been calculated for each segment in the model, and thermal-to-fast neutron flux ratios were determined for each component segment. Axial profiles of the spectra are provided for all components of the reactor. Individual segment statistical uncertainties were limited wherever possible, and the group fluxes for all important reflector components have a standard deviation below 10%.

  10. Heat-Flux Sensor For Hot Engine Cylinders

    NASA Technical Reports Server (NTRS)

    Kim, Walter S.; Barrows, Richard F.; Smith, Floyd A.; Koch, John

    1989-01-01

    Heat-flux sensor includes buried wire thermocouple and thin-film surface thermocouple, made of platinum and platinum with 13 percent rhodium. Sensor intended for use in ceramic-insulated, low-heat-rejection diesel engine at temperatures of about 1,000 K. Thermocouple junction resists environment in cylinder of advanced high-temperature diesel engine created by depositing overlapping films of Pt and 0.87 Pt/0.13 Rh on iron plug. Plug also contains internal thermocouple.

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. Results from tests on a high work transonic turbine for an energy efficient engine

    NASA Technical Reports Server (NTRS)

    Crow, D. E.; Welna, H.; Singer, I. D.; Vanco, M. R.

    1980-01-01

    The experimental results of the evaluation of two high work, transonic, single-stage turbines investigated under the Energy Efficient Engine (E3) Program are presented. The objective of the E3 program is to provide an advanced technology base for a new generation of fuel-conservative turbofan engines. A single-stage turbine required fewer cooled airfoils, a reduced number of leakage paths and no interstage seals. These advanced energy efficient engines require high engine pressure ratios resulting in high expansion ratio, transonic, turbine designs which must have high aerodynamic efficiency. The goal of the turbine program is to develop a high pressure turbine that is compatible with the overall engine design and has an uncooled efficiency of 90.8 percent.

  16. 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…

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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 1000%C2%B0C 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.

  3. 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

  4. 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.

  5. Engineering the fusion reactor first wall

    SciTech Connect

    Wurden, Glen; Scott, Willms

    2008-01-01

    Recently the National Academy of Engineering published a set of Grand Challenges in Engineering in which the second item listed was entitled 'Provide energy from fusion'. Clearly a key component of this challenge is the science and technology associated with creating and maintaining burning plasmas. This is being vigorously addressed with both magnetic and inertial approaches with various experiments such as ITER and NIF. Considerably less attention is being given to another key component of this challenge, namely engineering the first wall that will contain the burning plasma. This is a daunting problem requiring technologies and materials that can not only survive, but also perform multiple essential functions in this extreme environment. These functions are (1) shield the remainder of the device from radiation. (2) convert of neutron energy to useful heat and (3) breed and extract tritium to maintain the reactor fuel supply. The first wall must not contaminate the plasma with impurities. It must be infused with cooling to maintain acceptable temperatures on plasma facing and structural components. It must not degrade. It must avoid excessive build-up of tritium on surfaces, and, if surface deposits do form, must be receptive to cleaning techniques. All these functions and constraints must be met while being subjected to nuclear and thermal radiation, particle bombardment, high magnetic fields, thermal cycling and occasional impingement of plasma on the surface. And, operating in a nuclear environment, the first wall must be fully maintainable by remotely-operated manipulators. Elements of the first wall challenge have been studied since the 1970' s both in the US and internationally. Considerable foundational work has been performed on plasma facing materials and breeding blanket/shield modules. Work has included neutronics, materials fabrication and joining, fluid flow, tritium breeding, tritium recovery and containment, energy conversion, materials damage and

  6. Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

    SciTech Connect

    Cartier, J.; Casoli, P.; Chappert, F.

    2013-07-01

    In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

  7. High performance inboard shield design for the compact TIBER-II test reactor: Appendix A-2

    SciTech Connect

    El-Guebaly, L.A.; Sviatoslavsky, I.N.

    1987-01-01

    The compactness of the TIBER-II reactor has placed a premium on the design of a high performance inboard shield to protect the inner legs of the toroidal field (TF) coils. The available space for shield is constrained to 48 cm and the use of tungsten is mandatory to protect the magnet against the 1.53 MW/m/sup 2/ neutron wall loading. The primary requirement for the shield is to limit the fast neutron fluence to 10/sup 19/ n/cm/sup 2/. In an optimization study, the performance of various candidate materials for protecting the magnet was examined. The optimum shield consists of a 40 cm thick W layer, followed by an 8 cm thick H/sub 2/O/LiNO/sub 3/ layer. The mechanical design of the shield calls for tungsten blocks within SS stiffened panels. All the coolant channels are vertical with more of them in the front where there is a high heat load. The coolant pressure is 0.2 MPa and the maximum structural surface temperature is <95/sup 0/C. The effects of the detailed mechanical design of the shield and the assembly gaps between the shield sectors on the damage in the magnet were analyzed and peaking factors of approx.2 were found at the hot spots. 2 refs., 6 figs., 2 tabs.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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. ive reactors containing 66 L of slurry (30% soil by weight), were operated in parallel. he soil was a sandy soil with minor gravel content. he pilot-scale phase utilized an inoculum of...

  13. 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...

  14. Fast Flux Test Facility replacement of a primary sodium pump

    SciTech Connect

    Krieg, S.A.; Thomson, J.D.

    1985-11-15

    The Fast Flux Test Facility is a 400 MW Thermal Sodium Cooled Fast Reactor operated by Westinghouse Hanford Company for the US Department of Energy. During startup testing in 1979, the sodium level in one of the primary sodium pumps was inadvertently raised above the normal height. This resulted in distortion of the pump shaft. Pump replacement was carried out using special maintenance equipment. Nuclear radiation and contamination were not significant problems since replacement operations were carried out shortly after startup of the Fast Flux Test Facility.

  15. Durability testing of the AJ10-221 490 N high performance (321 sec Isp) engine

    NASA Technical Reports Server (NTRS)

    Jassowski, D. M.; Rosenberg, S. D.; Schoenman, L.

    1993-01-01

    The durability of the 490 N AJ10-221 engine is characterized on the basis of data from 93 tests and a total firing life of 6.3 hr. For the three Ir/Re chambers tested, no limiting conditions were encountered in the 43,379 sec and 229 test thermal cycles. A wide range of off nominal operating conditions was successfully demonstrated.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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

  1. 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.

  2. Measurement of local high-level, transient surface heat flux

    NASA Technical Reports Server (NTRS)

    Liebert, Curt H.

    1988-01-01

    This study is part of a continuing investigation to develop methods for measuring local transient surface heat flux. A method is presented for simultaneous measurements of dual heat fluxes at a surface location by considering the heat flux as a separate function of heat stored and heat conducted within a heat flux gage. Surface heat flux information is obtained from transient temperature measurements taken at points within the gage. Heat flux was determined over a range of 4 to 22 MW/sq m. It was concluded that the method is feasible. Possible applications are for heat flux measurements on the turbine blade surfaces of space shuttle main engine turbopumps and on the component surfaces of rocket and advanced gas turbine engines and for testing sensors in heat flux gage calibrators.

  3. High heat flux measurements and experimental calibrations/characterizations

    NASA Technical Reports Server (NTRS)

    Kidd, Carl T.

    1992-01-01

    Recent progress in techniques employed in the measurement of very high heat-transfer rates in reentry-type facilities at the Arnold Engineering Development Center (AEDC) is described. These advances include thermal analyses applied to transducer concepts used to make these measurements; improved heat-flux sensor fabrication methods, equipment, and procedures for determining the experimental time response of individual sensors; performance of absolute heat-flux calibrations at levels above 2,000 Btu/cu ft-sec (2.27 kW/cu cm); and innovative methods of performing in-situ run-to-run characterizations of heat-flux probes installed in the test facility. Graphical illustrations of the results of extensive thermal analyses of the null-point calorimeter and coaxial surface thermocouple concepts with application to measurements in aerothermal test environments are presented. Results of time response experiments and absolute calibrations of null-point calorimeters and coaxial thermocouples performed in the laboratory at intermediate to high heat-flux levels are shown. Typical AEDC high-enthalpy arc heater heat-flux data recently obtained with a Calspan-fabricated null-point probe model are included.

  4. 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.

  5. 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.

  6. The Advanced Test Reactor Irradiation Facilities and Capabilities

    SciTech Connect

    S. Blaine Grover; Raymond V. Furstenau

    2007-03-01

    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 different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The ATR has enhanced capabilities in experiment monitoring and control systems for instrumented and/or temperature controlled experiments. The control systems utilize feedback from thermocouples in the experiment to provide a custom blended flowing inert gas mixture to control the temperature in the experiments. Monitoring systems have also been utilized on the exhaust gas lines from the experiment to monitor different parameters, such as fission gases for fuel experiments, during irradiation. ATR’s unique control system provides axial flux profiles in the experiments, unperturbed by axially positioned control components, throughout each reactor operating cycle and over the duration of test programs requiring many years of irradiation. The ATR irradiation positions vary in diameter from 1.6 cm (0.625 inches) to 12.7 cm (5.0 inches) over an active core length of 122 cm (48.0 inches). Thermal and fast neutron fluxes can be adjusted radially across the core depending on the needs of individual test programs. This paper will discuss the different irradiation capabilities available and the cost/benefit issues related to each capability. Examples of different experiments will also be discussed to demonstrate the use of the capabilities and facilities at ATR for performing irradiation experiments.

  7. Eugene Wigner, The First Nuclear Reactor Engineer

    NASA Astrophysics Data System (ADS)

    Weinberg, Alvin M.

    2002-04-01

    All physicists recognize Eugene Wigner as a theoretical physicist of the very first rank. Yet Wigner's only advanced degree was in Chemical Engineering. His physics was largely self-taught. During WWII, Wigner brilliantly returned to his original occupation as an engineer. He led the small team of theoretical physicists and engineers who designed, in remarkable detail, the original graphite-moderated, water-cooled Hanford reactor, which produced the Pu239 of the Trinity and Nagasaki bombs. With his unparalleled understanding of chain reactors (matched only by Fermi) and his skill and liking for engineering, Wigner can properly be called the Founder of Nuclear Engineering. The evidence for this is demonstrated by a summary of his 37 Patents on various chain reacting systems.

  8. 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.

  9. 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.

  10. 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.

  11. High flux compact neutron generators

    SciTech Connect

    Reijonen, J.; Lou, T.-P.; Tolmachoff, B.; Leung, K.-N.; Verbeke, J.; Vujic, J.

    2001-06-15

    Compact high flux neutron generators are developed at the Lawrence Berkeley National Laboratory. The neutron production is based on D-D or D-T reaction. The deuterium or tritium ions are produced from plasma using either a 2 MHz or 13.56 MHz radio frequency (RF) discharge. RF-discharge yields high fraction of atomic species in the beam which enables higher neutron output. In the first tube design, the ion beam is formed using a multiple hole accelerator column. The beam is accelerated to energy of 80 keV by means of a three-electrode extraction system. The ion beam then impinges on a titanium target where either the 2.4 MeV D-D or 14 MeV D-T neutrons are generated. The MCNP computation code has predicted a neutron flux of {approximately}10{sup 11} n/s for the D-D reaction at beam intensity of 1.5 A at 150 kV. The neutron flux measurements of this tube design will be presented. Recently new compact high flux tubes are being developed which can be used for various applications. These tubes also utilize RF-discharge for plasma generation. The design of these tubes and the first measurements will be discussed in this presentation.

  12. Conversion of research and test reactors : status and current plans.

    SciTech Connect

    Roglans, J.; Staples, P.; Butler, N.; Nuclear Engineering Division

    2007-01-01

    reactors desired higher specific power and neutron flux and, to avoid costs associated with the development of higher density LEU fuels, those reactors used high-enriched uranium (HEU) material, with an enrichment of 20% or higher, and typically over 90%, with the existing fuel designs. As HEU fuel became readily available, it turned into the usual fuel for research and test reactors, even for some that had initially operated with LEU fuel.

  13. 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)

  14. Testing the applicability of neural networks as a gap-filling method using CH4 flux data from high latitude wetlands

    NASA Astrophysics Data System (ADS)

    Dengel, S.; Zona, D.; Sachs, T.; Aurela, M.; Jammet, M.; Parmentier, F. J. W.; Oechel, W.; Vesala, T.

    2013-12-01

    Since the advancement in CH4 gas analyser technology and its applicability to eddy covariance flux measurements, monitoring of CH4 emissions is becoming more widespread. In order to accurately determine the greenhouse gas balance, high quality gap-free data is required. Currently there is still no consensus on CH4 gap-filling methods, and methods applied are still study-dependent and often carried out on low resolution, daily data. In the current study, we applied artificial neural networks to six distinctively different CH4 time series from high latitudes, explain the method and test its functionality. We discuss the applicability of neural networks in CH4 flux studies, the advantages and disadvantages of this method, and what information we were able to extract from such models. Three different approaches were tested by including drivers such as air and soil temperature, barometric air pressure, solar radiation, wind direction (indicator of source location) and in addition the lagged effect of water table depth and precipitation. In keeping with the principle of parsimony, we included up to five of these variables traditionally measured at CH4 flux measurement sites. Fuzzy sets were included representing the seasonal change and time of day. High Pearson correlation coefficients (r) of up to 0.97 achieved in the final analysis are indicative for the high performance of neural networks and their applicability as a gap-filling method for CH4 flux data time series. This novel approach which we show to be appropriate for CH4 fluxes is a step towards standardising CH4 gap-filling protocols.

  15. Fast Flux Test Facility fuel and test management: The first 10 years

    SciTech Connect

    Bennett, R.A.; Bennett, C.L.; Campbell, L.R.; Dobbin, K.D.; Tang, E.L.

    1991-07-01

    Core design and fuel and test management have been performed efficiently at the Fast Flux Test Facility. No outages have been extended to adjust core loadings. Development of mixed oxide fuels for advanced liquid metal breeder reactors has been carried out successfully. In fact, the fuel performance is extraordinary. Failures have been so infrequent that further development and refinement of fuel requirements seem appropriate and could lead to a significant reduction in projected electrical busbar costs. The Fast Flux Test Facility is also involved in early metal fuel development tests and appears to be an ideal test bed for any further fuel development or refinement testing. 3 refs., 4 figs., 2 tabs.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Preserving physics knowledge at the fast flux test facility

    SciTech Connect

    Wootan, D.; Omberg, R.; Makenas, B. J.; Polzin, D. L.

    2012-07-01

    One of the goals of the Dept. 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. A disciplined and orderly approach has been developed 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. (authors)

  1. 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.

  2. Test reactor irradiation coordination

    SciTech Connect

    Heartherly, D.W.; Siman Tov, I.I.; Sparks, D.W.

    1995-10-01

    This task was established to supply and coordinate irradiation services needed by NRC contractors other than ORNL. These services include the design and assembly of irradiation capsules as well as arranging for their exposure, disassembly, and return of specimens. During this period, the final design of the facility and specimen baskets was determined through an iterative process involving the designers and thermal analysts. The resulting design should permit the irradiation of all test specimens to within 5{degrees}C of their desired temperature. Detailing of all parts is ongoing and should be completed during the next reporting period. Procurement of the facility will also be initiated during the next review period.

  3. Fast Flux Test Facility Closure Project - Project Management Plan

    SciTech Connect

    BEACH, R.R.

    2002-09-26

    The Fast Flux Test Facility (FFTF) Closure Project, Project Management Plan, Revision 5, provides the scope, cost, and schedule to achieve the most cost effective and expeditious closure of the FFTF to an assumed final end-state with the reactor vessel and the containment building, below the 5504 grade level, being entombed in place. Closure will be completed by December 2009 at a cost of $547 million.

  4. 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).

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

    PubMed

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

    2015-01-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-10-01

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

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

    PubMed Central

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

    2015-01-01

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

  8. 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

  9. Testing the applicability of neural networks as a gap-filling method using CH4 flux data from high latitude wetlands

    NASA Astrophysics Data System (ADS)

    Dengel, S.; Zona, D.; Sachs, T.; Aurela, M.; Jammet, M.; Parmentier, F. J. W.; Oechel, W.; Vesala, T.

    2013-05-01

    Since the advancement in CH4 gas analyser technology and its applicability to eddy covariance flux measurements, monitoring of CH4 emissions is becoming more widespread. In order to accurately determine the greenhouse gas balance, high quality gap-free data is required. Currently there is still no consensus on CH4 gap-filling methods, and methods applied are still study-dependent and often carried out on low resolution daily data. In the current study, we applied artificial neural networks to six distinctively different CH4 time series from high latitudes in order to recover missing data points, explained the method and tested its functionality. We discuss the applicability of neural networks in CH4 flux studies, the advantages and disadvantages of this method, and what information we were able to extract from such models. In keeping with the principle of parsimony, we included only five standard meteorological variables traditionally measured at CH4 flux measurement sites. These included drivers such as air and soil temperature, barometric air pressure, solar radiation, and in addition wind direction (indicator of source location). Four fuzzy sets were included representing the time of day. High Pearson correlation coefficients (r) of 0.76-0.93 achieved in the final analysis are indicative for the high performance of neural networks and their applicability as a gap-filling method for CH4 flux data time series. This novel approach that we showed to be appropriate for CH4 fluxes is a step towards standardising CH4 gap-filling protocols.

  10. 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.

  11. 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.

  12. 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.

  13. INITIAL IRRADIATION OF THE FIRST ADVANCED GAS REACTOR FUEL DEVELOPMENT AND QUALIFICATION EXPERIMENT IN THE ADVANCED TEST REACTOR

    SciTech Connect

    S. Blaine Grover; David A. Petti

    2007-09-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The 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 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 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 for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  14. 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.

  15. 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.

  16. 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

  17. Shear punch testing of candidate reactor materials after irradiation in fast reactors and spallation environments

    NASA Astrophysics Data System (ADS)

    Maloy, S. A.; Romero, T. J.; Hosemann, P.; Toloczko, M. B.; Dai, Y.

    2011-10-01

    Ferritic/martensitic steels and nickel-base superalloys are potential materials for use in spallation targets and fusion and fast reactors. To investigate the effects of irradiation on these materials, tests were performed after irradiation in the high energy proton beam at the Paul Scherrer Institute (SINQ Target Irradiation Program (STIP), 570 MeV), as well on specimens obtained from a driver duct irradiated in the Fast Flux Test Facility (FFTF). Dose accumulations were up to 18 dpa for STIP irradiations (at 147-406 °C) and up to 155 dpa in FFTF (at 383-505 °C). The helium/dpa ratios ranged from 0.2 to 80 appm/dpa. Mechanical testing was performed at 25 °C. Increases in shear yield and shear maximum stress with increasing dose mirrored the results observed from companion tensile tests.

  18. 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.

  19. 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.

  20. 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.

  1. High-Flux and Low-Flux Membranes: Efficacy in Hemodialysis

    PubMed Central

    Oshvandi, Khodayar; Kavyannejad, Rasol; Borzuo, Sayed Reza; Gholyaf, Mahmoud

    2014-01-01

    Background: Inadequacy of dialysis is one of the main causes of death in hemodialysis patients. Some studies have suggested that high‐flux membrane improves the removal of moderate-sized molecules while other studies indicate no significant effect on them. Objectives: This study aimed to investigate the dialysis efficacy of low-flux versus high-flux membranes in hemodialysis patients. Patients and Methods: Forty hemodialysis patients participated in this cross-over clinical trial. Two sessions of low-flux and high-flux membrane dialysis were performed consecutively, in the first and second stage of the trial. In both stages, blood samples before and after the dialysis were taken and sent to the laboratory for assessment. Blood urea nitrogen (BUN), KT/V and the urea reduction ratio (URR) indexes were used to determine dialysis efficacy. Data were analyzed using t test and paired t test. Results: The mean KT/V was 1.27 ± 0.28 in high-flux and 1.10 ± 0.32 in low-flux membrane which, these differences were statistically significant (P = 0.017). The mean of URR was 0.65 ± 0.09 in high-flux and 0.61 ± 0.14 in low-flux membrane, which these differences were not statistically significant (P = 0.221). Conclusions: The high-flux membrane had better dialysis adequacy, so we suggest using high-flux membrane in hemodialysis centers. PMID:25699283

  2. A high heat flux experiment for verification of thermostructural analysis

    NASA Technical Reports Server (NTRS)

    Gladden, Herbert J.; Melis, Matthew E.

    1988-01-01

    A major concern in advancing the state of the art technologies for hypersonic vehicles is the development of an aeropropulsion system capable of handling the high heat fluxes during flight. The leading edges of such systems must not only tolerate the maximum heating rates, but must also minimize distortions to the flow field due to excessive blunting and/or thermal warping of the compression surface to achieve the high inlet performance required. A combined analytical and experimental effort to study the aerothermodynamic loads on actively cooled structures for hypersonic applications was established. A hydrogen/oxygen rocket engine was modified to establish a high enthalpy high heat flux environment. The facility provides heat flux levels from about 200 up to 10000 Btu/sq ft/sec. Cross flow and parallel flow regeneratively cooled model can be tested and analyzed by using cooling fluids of water and hydrogen. Results are presented of the experiment and the characteristics of the Hot Gas Test Facility. The predicted temperature results of the cross flow model are compared with the experimental data on the first monolithic specimens and are found to be in good agreement. Thermal stress analysis results are also presented.

  3. 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.

  4. 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.

  5. Knowledge Preservation at the Fast Flux Test Facility

    SciTech Connect

    Wootan, David W.; Omberg, Ronald P.

    2011-12-30

    One of the goals of the Department of Energy's Office of Nuclear Energy Fuel Cycle Research and Development Program (FCRD) 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. 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. A disciplined and orderly approach has been developed 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.

  6. 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.

  7. 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.

  8. Prevention of significant deterioration application for approval to construct SP-100 Ground Engineering System Test Site

    SciTech Connect

    Not Available

    1990-04-01

    The following application is being submitted by the US Department of Energy, Richland Operations Office, P.O. Box 550, Richland, Washington 99352, pursuant to WAC 173-403-080, and in compliance with the Department of Ecology Guide to Processing a Prevention of Significant Deterioration (PSD) Permit'' for a new source of airborne radionuclide emissions at the Hanford Site in Washington State. The new source, the SP-100 Ground Engineering System (GES) Test Site, will be located in the 309 Building of the 300 Area. The US Department of Energy (DOE), the National Aeronautics and Space Administration (NASA), and the US Department of Defense (DOD) have entered into an agreement to jointly develop space nuclear reactor power system technology. The DOE has primary responsibility for developing and ground testing the nuclear subsystem. A ground test of a reactor is necessary to demonstrate technology readiness of this major subsystem before proceeding with the flight system development and demonstration. The SP-100 GES Test Site will provide a location for the operation and testing of a prototype space-based, liquid metal-cooled, fast flux nuclear reactor in an environment closely simulating the vacuum and temperature conditions of space operations. The purpose of the GES is to develop safe, compact, light-weight and durable space reactor power system technology. This technology will be used to provide electric power, in the range of tens to hundreds of kilowatts, for a variety of potential future civilian and military space missions requiring long-term, high-power level sources of energy. 20 refs., 8 figs., 7 tabs.

  9. 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

  10. 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

  11. Control of Tritium at the Fast Flux Test Facility

    SciTech Connect

    Prevo, P. R.

    1989-06-01

    The Fast Flux Test Facility (FFTF) is located on the Hanford Site near Richland, Washington, and is operated by the Westinghouse Hanford Company for the US Department of Energy (DOE). The facility features a 400 MW(t) three-loop sodium-cooled, mixed-oxide-fueled reactor that was designed for irradiation testing of fuels and materials to support the commercial development of liquid metal fast reactors. The mission has subsequently been expanded to include a passive safety test program, irradiation of fusion and space reactor materials, and isotope production. The FFTF has been in operation for about 7 yr, which includes over 1600 d of full power operation of the fast test reactor (FTR). Radiological operating experience at the FFTF has been excellent. Collective dose equivalents received by operating personnel have been very low (5 person-rem/yr average). No major contamination problems have been encountered in operating and maintaining the plant, and release of radioactivity to the environment has been well below acceptable limits (Bunch and Prevo 1987). Skin contamination events have averaged less than two per year. There have been no internal depositions. This paper discusses the generation, transport and distribution, and radiological aspects of tritium control at the FFTF. 3 refs., 1 fig., 2 tabs.

  12. 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.

  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. 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…

  15. 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...

  16. 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.

  17. Italian High-speed Airplane Engines

    NASA Technical Reports Server (NTRS)

    Bona, C F

    1940-01-01

    This paper presents an account of Italian high-speed engine designs. The tests were performed on the Fiat AS6 engine, and all components of that engine are discussed from cylinders to superchargers as well as the test set-up. The results of the bench tests are given along with the performance of the engines in various races.

  18. 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.

  19. 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.

  20. 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)

  1. 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.

  2. Benchmarking transition costs for the Fast Flux Test Facility

    SciTech Connect

    Hulvey, R.K.

    1996-12-31

    The Fast Flux Test Facility (FFTF) is a government-owned, 400-MW(thermal), sodium-cooled test reactor operated by Westinghouse Hanford Company. The reactor is shut down and is undergoing a transition to a long-term surveillance and maintenance state. The mission strategy for the FFTF transition project is to place the FFTF in a radiologically and industrially safe condition, completing the transition phase activities as soon as possible to drive down the current annual surveillance and maintenance costs from approximately $26 million/yr to roughly $1.5 million/yr. The effort to establish the shutdown and transition costs for this 7-yr, $260 million activity is a first of a kind for the U.S. Department of Energy (DOE).

  3. High performance jet-engine flight test data base for HSR

    NASA Astrophysics Data System (ADS)

    Kelly, Jeffrey

    1992-04-01

    The primary acoustic priority of the flight test data base for HSR is the validation of the NASA Aircraft Noise Prediction Program (ANOPP) and other source noise codes. Also, the noise measurements are an important support function for the High Lift Program devoted to HSR. Another concern that will be addressed is a possible noise problem 7-20 miles from take-off during climbout. The attention arises from the higher speeds envisioned for the HSCT compared to conventional aircraft causing levels to increase because of Doppler amplification in conjunction with high source levels due to jet noise. An attempt may be made to measure airframe noise for the F-16XL test which would provide an assessment of this noise component for delta wing aircraft.

  4. High performance jet-engine flight test data base for HSR

    NASA Technical Reports Server (NTRS)

    Kelly, Jeffrey

    1992-01-01

    The primary acoustic priority of the flight test data base for HSR is the validation of the NASA Aircraft Noise Prediction Program (ANOPP) and other source noise codes. Also, the noise measurements are an important support function for the High Lift Program devoted to HSR. Another concern that will be addressed is a possible noise problem 7-20 miles from take-off during climbout. The attention arises from the higher speeds envisioned for the HSCT compared to conventional aircraft causing levels to increase because of Doppler amplification in conjunction with high source levels due to jet noise. An attempt may be made to measure airframe noise for the F-16XL test which would provide an assessment of this noise component for delta wing aircraft.

  5. 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.

  6. Space reactor fuel element testing in upgraded TREAT

    SciTech Connect

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

    1993-01-15

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is 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 suggests that full-scale PBR elements could be tested at an average energy deposition of [similar to]60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of [similar to]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. 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.

  9. 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.

  10. Preserving Physics Knowledge at the Fast Flux Test Facility

    SciTech Connect

    Wootan, David W.; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.; Nelson, Joseph V.; Polzin, David L.

    2011-11-01

    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. A disciplined and orderly approach has been developed 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.

  11. 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.

  12. A New High-Flux Chemical and Materials Crystallography Station at the SRS Daresbury. 1. Design, Construction and Test Results.

    PubMed

    Cernik, R J; Clegg, W; Catlow, C R; Bushnell-Wye, G; Flaherty, J V; Greaves, G N; Burrows, I; Taylor, D J; Teat, S J; Hamichi, M

    1997-09-01

    A new single-crystal diffraction facility has been constructed on beamline 9 of the SRS at Daresbury Laboratory for the study of structural problems in chemistry and materials science. The station utilizes up to 3.8 mrad horizontally from the 5 T wiggler magnet which can be focused horizontally and vertically. The horizontal focusing is provided by a choice of gallium-cooled triangular bent Si (111) or Si (220) monochromators, giving a wavelength range from 0.3 to 1.5 A. Focusing in the vertical plane is achieved by a cylindrically bent zerodur mirror with a 300 mum-thick palladium coating. The station is equipped with a modified Enraf-Nonius CAD-4 four-circle diffractometer and a Siemens SMART CCD area-detector system. High- and low-temperature facilities are available to cover the temperature range from about 80 to 1000 K. Early results on test compounds without optimization of the beam optics demonstrate that excellent refined structures can be obtained from samples giving diffraction patterns too weak to be measured with conventional laboratory X-ray sources, fulfilling a major objective of the project. PMID:16699241

  13. 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.

  14. 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.

  15. 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)

  16. Coil Designs for Novel Magnetic Geometries to Cure the Divertor Heat Flux Problem for Reactors

    NASA Astrophysics Data System (ADS)

    Pekker, M.; Valanju, P.; Kotschenreuther, M.; Wiley, J. C.; Strickler, D.

    2004-11-01

    Coil designs are developed for novel magnetic divertor geometries with a second axi-symmetric x-point and flux expansion region along the separatrix. Adjacent posters describe how these lead to spreading of heat flux and the possibility of stable, complete detachment to overcome serious physics and engineering problems in reactors. The principal feasibility issue is creating, with simple coils, additional X-points on the separatrix without extensively deforming the magnetic field in the main plasma. For the spherical tokamak NSTX, we show that adding one or two poloidal coils suffices to create a divergent flux at the divertor, i.e., a new x-point. The currents and forces for the extra coils are small. We also modify ARIES ST design to show reactor feasibility. Optimized coil designs for PEGASUS, ARIES RS/AT, and a modular ITER retrofit are also being developed. For our calculations we used self consistent code FBEQ, which was used to design NSTX. We also use NCSX tools for optimization of designs with competing physics and engineering constraints.

  17. Securing Operating Data From Passive Safety Tests at the Fast Flux Test Facility

    SciTech Connect

    Wootan, David W.; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.; Nelson, Joseph V.; Polzin, David L.

    2011-06-01

    The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to operate in the United States, from 1982 to 1992. The technologies employed in designing and constructing this reactor, along with information obtained from tests conducted during its operation, are currently being secured and archived by the Department of Energy’s Office of Nuclear Energy Fuel Cycle Research and Development Program. This report is one in a series documenting the overall project efforts to retrieve and preserve critical information related to advanced reactors. A previous report summarized the initial efforts to review, retrieve and preserve the most salient documents related to Passive Safety Testing (PST) in the FFTF. Efforts continue to locate, secure, and retrieve record copies of original plant data tapes for the series of passive safety tests conducted between 1986 and 1991.

  18. Securing Operating Data From Passive Safety Tests at the Fast Flux Test Facility

    SciTech Connect

    Wootan, David W.; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.; Nelson, Joseph V.; Polzin, David L.

    2011-06-01

    The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to operate in the United States, from 1982 to 1992. The technologies employed in designing and constructing this reactor, along with information obtained from tests conducted during its operation, are currently being secured and archived by the Department of Energy's Office of Nuclear Energy Fuel Cycle Research and Development Program. This report is one in a series documenting the overall project efforts to retrieve and preserve critical information related to advanced reactors. A previous report summarized the initial efforts to review, retrieve and preserve the most salient documents related to Passive Safety Testing (PST) in the FFTF. Efforts continue to locate, secure, and retrieve record copies of original plant data tapes for the series of passive safety tests conducted between 1986 and 1991.

  19. 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.

  20. Fast flux test facility radioisotope production and medical applications

    SciTech Connect

    Schenter, R.E.; Smith, S.G.; Tenforde, T.S.

    1997-12-01

    The Fast Flux Test Facility (FFTF) is a 400-MW, sodium-cooled reactor that operated successfully from 1982 to 1992, conducting work in support of the liquid-metal reactor industry by developing and testing fuel assemblies, control rods, and other core reactor components. Upon termination of this program, the primary mission of FFTF ended, and it was placed in a standby mode in 1993. However, in January 1997 the U.S. Secretary of Energy requested that FFTF be evaluated for a future mission that would consist of a primary goal of producing tritium for nuclear defense applications and a secondary goal of supplying medical isotopes for research and clinical applications. Production by FFTF of tritium for U.S. nuclear weapons would augment the dual-track strategy now under consideration for providing a long-term tritium supply in the United States (consisting of a light water reactor option and an accelerator option). A decision by the Secretary of Energy on proceeding with steps leading toward the possible reactivation of FFTF will be made before the end of 1998.

  1. Ultra high temperature particle bed reactor design

    SciTech Connect

    Lazareth, O.; Ludewig, H.; Perkins, K.; Powell, J.

    1990-01-01

    This study is a computer analysis of a conceptual nuclear reactor. The purpose of this work is to design a direct nuclear propulsion engine which could be used for a mission to Mars. The main features of this reactor design are high values for I{sub sp} and, secondly, 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 H{sub 2} 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.

  2. High-temperature, high-pressure testing of zinc titanate in a bench-scale fluidized-bed reactor for 100 cycles

    SciTech Connect

    Gupta, R.P.; Gangwal, S.K.

    1993-06-01

    Integrated gasification combined cycle (IGCC) power plants are being advanced worldwide to produce electricity from coal owing to their potential for superior environmental performance, economics, and efficiency in comparison to conventional coal-based power plants. A key component of these plants is a hot-gas desulfurization system employing efficient regenerable mixed-metal oxide sorbents. Leading sorbent candidates include zinc ferrite and zinc titanate. These sorbents can remove hydrogen sulfide (H{sub 2}S) in the fuel gas down to very low levels (typically <20 ppmv) at 500 to 750{degree}C and can be readily regenerated for multicycle operation with air. To this end, the Research Triangle Institute (RTI) has formulated and tested a series of zinc titanate sorbents in a high-temperature, high- pressure HTHP fluidized-bed bench-scale reactor. Multicycle HTHP bench-scale testing of these sorbents under a variety of conditions culminated in the development of a ZT-4 sorbent that exhibited the best overall performance in terms of chemical reactivity, sulfur capacity, regenerability, structural properties, and attrition resistance. Following this parametric study, a life-cycle test consisting of 100 sulfidation-regeneration cycles was carried out with ZT-4 in the bench unit.

  3. 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)

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. Reactor calculation benchmark PCA blind test results

    SciTech Connect

    Kam, F.B.K.; Stallmann, F.W.

    1980-01-01

    Further improvement in calculational procedures or a combination of calculations and measurements is necessary to attain 10 to 15% (1 sigma) accuracy for neutron exposure parameters (flux greater than 0.1 MeV, flux greater than 1.0 MeV, and dpa). The calculational modeling of power reactors should be benchmarked in an actual LWR plant to provide final uncertainty estimates for end-of-life predictions and limitations for plant operations. 26 references, 14 figures, 6 tables.

  9. Maintenance Implementation Plan for the Fast Flux Test Facility

    SciTech Connect

    Crawford, C.N.; Duffield, M.F.

    1992-06-01

    The maintenance program for the 400 Area, Fast Flux Test Facility (FFTF)Plant and plant support facilities includes the reactor plant, reactor support systems and equipment, Maintenance and Storage Facility, plant buildings, and building support systems. These are the areas of the facility that are covered by this plan. The personnel support facilities and buildings are maintained and supported by another department within Westinghouse Hanford, and are not included here. The FFTF maintenance program conducts the corrective and preventive maintenance necessary to ensure the operational reliability and safety of the reactor plant and support equipment. This comprehensive maintenance program also provides for maximizing the useful life of plant equipment and systems to realize the most efficient possible use of resources. The long-term future of the FFTF is uncertain; in the near term, the facility is being placed in standby. As the plant transitions from operating status to standby, the scope of the maintenance program will change from one of reactor operational reliability and life extension to preservation.

  10. Thin film heat flux sensor for Space Shuttle Main Engine turbine environment

    NASA Technical Reports Server (NTRS)

    Will, Herbert

    1991-01-01

    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.

  11. 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

  12. 96. ARAIV. Typical view of ML1 reactor field test operations. ...

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

    96. ARA-IV. Typical view of ML-1 reactor field test operations. Reactor has been hoisted onto truck at door of test assembly building. May 17, 1963. Ineel photo no. 63-3257. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

  13. 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.

  14. 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.

  15. Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2006-10-01

    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-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The 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 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. 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 for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  16. 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.

  17. 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.

  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. PMID:24593452

  19. 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.

  20. 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. )

    1992-01-01

    On December 2, 1980, the Fast Flux Test Facility (FFTF) reached its full design power of 400 MW for the first time. From the start, the FFTF provided a modern liquid-metal reactor (LMR) test facility recognized for excellence, innovation, and efficiency of operation. Its unique instrumentation and special test capabilities have allowed the facility to stay at the cutting edge of technology. Prototypical size and core environment allow the FFTF to demonstrate core components and directly support design optimization of LMRs. Since December 1980, the FFTF has irradiated > 64,000 mixed-oxide driver and test fuel pins, > 1,000 metal-fueled pins, > 100 carbide-fueled pins, and > 35 nitride-fueled pins (supporting the U.S. space reactor program). This paper reviews the status of one of the major activities at the FFTF for its first 12 yr of operation - DOE-sponsored testing and development of fuel, blanket, and absorber assemblies for commercial LMRs.

  1. Correlation of the Characteristics of Single-Cylinder and Flight Engines in Tests of High-Performance Fuels in an Air-Cooled Engine I : Cooling Characteristics

    NASA Technical Reports Server (NTRS)

    Wilson, Robert W.; Richard, Paul H.; Brown, Kenneth D.

    1945-01-01

    Variable charge-air flow, cooling-air pressure drop, and fuel-air ration investigations were conducted to determine the cooling characteristics of a full-scale air-cooled single cylinder on a CUE setup. The data are compared with similar data that were available for the same model multicylinder engine tested in flight in a four-engine airplane. The cylinder-head cooling correlations were the same for both the single-cylinder and the flight engine. The cooling correlations for the barrels differed slightly in that the barrel of the single-cylinder engine runs cooler than the barrel of te flight engine for the same head temperatures and engine conditions.

  2. 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.

  3. 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.

  4. 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.

  5. Cold flow testing of the Space Shuttle Main Engine alternate turbopump development high pressure fuel turbine model

    NASA Technical Reports Server (NTRS)

    Gaddis, Stephen W.; Hudson, Susan T.; Johnson, P. D.

    1992-01-01

    NASA's Marshall Space Flight Center has established a cold airflow turbine test program to experimentally determine the performance of liquid rocket engine turbopump drive turbines. Testing of the SSME alternate turbopump development (ATD) fuel turbine was conducted for back-to-back comparisons with the baseline SSME fuel turbine results obtained in the first quarter of 1991. Turbine performance, Reynolds number effects, and turbine diagnostics, such as stage reactions and exit swirl angles, were investigated at the turbine design point and at off-design conditions. The test data showed that the ATD fuel turbine test article was approximately 1.4 percent higher in efficiency and flowed 5.3 percent more than the baseline fuel turbine test article. This paper describes the method and results used to validate the ATD fuel turbine aerodynamic design. The results are being used to determine the ATD high pressure fuel turbopump (HPFTP) turbine performance over its operating range, anchor the SSME ATD steady-state performance model, and validate various prediction and design analyses.

  6. Cold flow testing of the Space Shuttle Main Engine alternate turbopump development high pressure fuel turbine model

    NASA Astrophysics Data System (ADS)

    Gaddis, Stephen W.; Hudson, Susan T.; Johnson, P. D.

    1992-06-01

    NASA's Marshall Space Flight Center has established a cold airflow turbine test program to experimentally determine the performance of liquid rocket engine turbopump drive turbines. Testing of the SSME alternate turbopump development (ATD) fuel turbine was conducted for back-to-back comparisons with the baseline SSME fuel turbine results obtained in the first quarter of 1991. Turbine performance, Reynolds number effects, and turbine diagnostics, such as stage reactions and exit swirl angles, were investigated at the turbine design point and at off-design conditions. The test data showed that the ATD fuel turbine test article was approximately 1.4 percent higher in efficiency and flowed 5.3 percent more than the baseline fuel turbine test article. This paper describes the method and results used to validate the ATD fuel turbine aerodynamic design. The results are being used to determine the ATD high pressure fuel turbopump (HPFTP) turbine performance over its operating range, anchor the SSME ATD steady-state performance model, and validate various prediction and design analyses.

  7. Flux attenuation at NREL's High-Flux Solar Furnace

    NASA Astrophysics Data System (ADS)

    Bingham, Carl E.; Scholl, Kent L.; Lewandowski, Allan A.

    1994-10-01

    The High-Flux Solar Furnace (HFSF) at the National Renewable Energy Laboratory (NREL) has a faceted primary concentrator and a long focal-length-to-diameter ratio (due to its off-axis design). Each primary facet can be aimed individually to produce different flux distributions at the target plane. Two different types of attenuators are used depending on the flux distribution. A sliding-plate attenuator is used primarily when the facets are aimed at the same target point. The alternate attenuator resembles a venetian blind. Both attenuators are located between the concentrator and the focal point. The venetian-blind attenuator is primarily used to control the levels of sunlight failing on a target when the primary concentrators are not focused to a single point. This paper will demonstrate the problem of using the sliding-plate attenuator with a faceted concentrator when the facets are not aimed at the same target point. We will show that although the alternate attenuator necessarily blocks a certain amount of incoming sunlight, even when fully open, it provides a more even attenuation of the flux for alternate aiming strategies.

  8. NATO Engine Test AT RRAD

    SciTech Connect

    Harry M. Meyer III

    2003-03-26

    This report details the reasons for and the outcome of a diesel engine test performed at the Red River Army Depot (RRAD) as part of a program called the Bradley Fighting Vehicle Component Reclamation through Thermal spray coating Technology Program.

  9. Nuclear rocket engine design based on the particle bed reactor

    SciTech Connect

    Powell, J.R.; Ludewig, H.; Mughabghab, S.; Lazareth, O. Jr.; Schmidt, E.; Maise, G. )

    1993-01-01

    A nuclear thermal rocket (NTR) engine design based on the particle-bed reactor (PBR) concept is described in this paper. This engine is designed to satisfy a mission to Mars and thus must develop a thrust of [approximately]1.75 (6) N. This requirement is satisfied if the reactor generates 2000 MW of power.

  10. 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.

  11. 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.

  12. 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.

  13. Cryogenic spin testing of NASA's shuttle engines

    NASA Astrophysics Data System (ADS)

    Maillar, Kenneth M.; Enos, Anthony; Gauthier, Robert

    1992-12-01

    Spin testing of the Space Shuttle Main Engine (SSME) high-pressure turbopump rotors is described focusing on the SSME cryogenic spin test facility. Testing at full operating speed is predicated on achieving and maintaining a cryogenic rotor temperature. Rotors are driven to operational speeds after being chilled to - 195 C.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. High Precision Mechanical Components for Soft X-ray Beamline: Engineering Goal and Testing Results

    SciTech Connect

    Kaznacheyev, K. V.; Karunakaran, Ch.; Sitnikov, A.; Loken, D.; Warwick, T.; Nagy, M.; Hitchcock, A. P.

    2007-01-19

    As the emittance of SR rings approaches the diffraction limit for soft x-rays, one requires not only excellence in design and performance of the optical elements, but also precision and performance of mechanical components, such as mirror manipulators, monochromator scanners and exit slits. We will present simple but efficient solutions for the mechanical systems of this type, commonly encountered in soft x-ray beamlines. These solutions have been implemented and their performance evaluated with test results from the spectromicroscopy beamline at the Canadian Light Source.

  19. High Precision Mechanical Components for Soft X-ray Beamline: Engineering Goal and Testing Results

    NASA Astrophysics Data System (ADS)

    Kaznacheyev, K. V.; Karunakaran, Ch.; Sitnikov, A.; Loken, D.; Warwick, T.; Nagy, M.; Hitchcock, A. P.

    2007-01-01

    As the emittance of SR rings approaches the diffraction limit for soft x-rays, one requires not only excellence in design and performance of the optical elements, but also precision and performance of mechanical components, such as mirror manipulators, monochromator scanners and exit slits. We will present simple but efficient solutions for the mechanical systems of this type, commonly encountered in soft x-ray beamlines. These solutions have been implemented and their performance evaluated with test results from the spectromicroscopy beamline at the Canadian Light Source.

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

    NASA Astrophysics Data System (ADS)

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

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

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. Testing of the United Stirling 4-95 solar Stirling engine on test bed concentrator

    NASA Technical Reports Server (NTRS)

    Nelving, H. G.

    1984-01-01

    The objectives with the testing, test set-ups, component designs, and the results of the testing of the solar Stirling engine in a parabolic dish system are presented. The most important tests are characterization of receivers, full day performance of complete system, cavity and aperture window test including influence from windeffects, control system tests, radiator system tests and special temperature measurements with infrared camera. The influence on performance of flux distribution depnding on concentrator alignment, and the optimum receiver operating criteria when balancing flux and temperatures on cooled receiver surface while avoiding flux on uncooled surfaces are also discussed.

  9. 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)

  10. 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.

  11. 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%.

  12. 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.

  13. 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).

  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. High Stability Engine Control (HISTEC)

    NASA Technical Reports Server (NTRS)

    DeLaat, John C.; Southwick, Robert D.; Gallops, George W.

    1996-01-01

    Future aircraft turbine engines, both commercial and military, must be able to successfully accommodate expected increased levels of steady-state and dynamic engine-face distortion. The current approach of incorporating a sufficient component design stall margin to tolerate these increased levels of distortion would significantly reduce performance. The objective of the High Stability Engine Control (HISTEC) program is to design, develop, and flight demonstrate an advanced, high-stability, integrated engine control system that uses measurement-based, real-time estimates of distortion to enhance engine stability. The resulting distortion tolerant control reduces the required design stall margin, with a corresponding increase in performance and decrease in fuel burn. The HISTEC concept, consisting of a Distortion Estimation System and a Stability Management Control, has been designed and developed. The Distortion Estimation System uses a small number of high-response pressure sensors at the engine face to calculate indicators of the type and extent of distortion in real time. The Stability Management Control, through direct control of the fan and compressor pressure ratio, accommodates the distortion by transiently increasing the amount of stall margin available based on information from the Distortion Estimation System. Simulation studies have shown the HISTEC distortion tolerant control is able to successfully estimate and accommodate time-varying distortion. Currently, hardware and software systems necessary for flight demonstration of the HISTEC concept are being designed and developed. The HISTEC concept will be flight tested in early 1997.

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

    PubMed

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

    2015-10-01

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

  17. Test data report, low speed wind tunnel tests of a full scale lift/cruise-fan inlet, with engine, at high angles of attack

    NASA Technical Reports Server (NTRS)

    Shain, W. M.

    1978-01-01

    A low speed wind tunnel test of a fixed lip inlet with engine, was performed. The inlet was close coupled to a Hamilton Standard 1.4 meter, variable pitch fan driven by a lycoming T55-L-11A engine. Tests were conducted with various combinations of inlet angle of attack freestream velocities, and fan airflows. Data were recorded to define the inlet airflow separation boundaries, performance characteristics, and fan blade stresses. The test model, installation, instrumentation, test, data reduction and final data are described.

  18. 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. PMID:25597686

  19. 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.

  20. 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.