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Sample records for high flux engineering test reactor

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

    SciTech Connect

    Guangquan, L.

    1988-01-01

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

  2. Neutron fluxes in test reactors

    SciTech Connect

    Youinou, Gilles Jean-Michel

    2017-01-01

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

  3. High flux reactor

    SciTech Connect

    Lake, J.A.; Heath, R.L.; Liebenthal, J.L.; DeBoisblanc, D.R.; Leyse, C.F.; Parsons, K.; Ryskamp, J.M.; Wadkins, R.P.; Harker, Y.D.; Fillmore, G.N.

    1988-08-16

    A high flux nuclear reactor is described comprising: (a) a pressure vessel including reactor coolant inlet means at the first end thereof and reactor coolant outlet means at the second end thereof; (b) a reactor coolant; (c) a first core segment housed within the pressure vessel, the first core segment including a plurality of concentric, circumferential fuel plates, the spacing between the concentric fuel plates forming coolant flow channels, each fuel plate being thin relative to the spacing between the fuel plates; (d) means for stationarily supporting the first core segment from the pressure vessel; (e) a second core segment housed within the pressure vessel and spaced axialy apart from the first core segment such that a coolant mixing plenum is formed therebetween, the second core segment including a plurality of concentric, circumferential fuel plates, the spacing between the concentric fuel plates forming coolant flow channels, each fuel plate being thin relative to the spacing between the fuel plates; (f) means for stationarily supporting the second core segment from the pressure vessel; and (g) first core coolant bypass means for channeling a volume of the coolant between the inlet means and the coolant mixing plenum such that the coolant volume bypasses the first core segment.

  4. 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.

  5. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2008-10-01

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

  7. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2011-03-01

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

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

    PubMed

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

    2011-03-01

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

  10. 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.

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

    SciTech Connect

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

    1992-08-01

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

  12. British high flux beam reactor.

    PubMed

    Egelstaff, P A

    1970-10-24

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

  13. 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.

  14. 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.

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

    NASA Technical Reports Server (NTRS)

    Melis, Matthew E.; Gladden, Herbert J.

    1990-01-01

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

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

    NASA Technical Reports Server (NTRS)

    Melis, Matthew E.; Gladden, Herbert J.

    1990-01-01

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

  17. 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.

  18. 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.

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

    SciTech Connect

    Jeffrey Bryan

    2009-06-01

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

  20. 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.

  1. 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)

  2. 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.

  3. UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

    SciTech Connect

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

    2016-03-11

    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 UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} 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 {sup 235}U in loaded fuel is 18.0% or above.

  4. Magnet technology for the Engineering Test Reactor

    SciTech Connect

    Henning, C.D.; Miller, J.R.

    1987-09-17

    The consideration for building an international Engineering Test Reactor emerged from the November 1985 Geneva Summit, in which President Reagan and Secretary Gorbachev called for the ''widest practical development of international cooperation'' in fusion. In parallel with the OTR design in the USSR, the FER in Japan, and the NET in Europe, the US has pursued the TIBER (Tokamak Ignition/Burn Experimental Reactor). This compact design of 3-m major radius is achievable because of high-current-density, radiation-tolerant magnets with nuclear heating rates up to 10 mW . cm/sup -3/. Full development of cable-in-conduit conductors (CICC) is seen as a credible path to achieve 40 A . mm/sup -2/ at fields of 12 T in the toroidal field (TF) coils and 14 T in the central poloidal field (PF) coils. Since neutron fluences of up to 10/sup 19/ n . cm/sup -2/ are expected in the TF coils, the unalloyed niobium-tin would be superior at 12 T. However, the central PF coil at 14 T is better shielded, so modified niobium tin would be advantageous. Polyimide insulation in the TF coils would withstand the equivalent 10/sup 10/) rads if loads in the winding pack are taken in compression. 13 refs., 7 figs., 2 tabs.

  5. 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

  6. 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.

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

    SciTech Connect

    1995-06-01

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

  8. 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.

  9. Calculation of heating values for the high flux isotope reactor

    SciTech Connect

    Peterson, J.; Ilas, G.

    2012-07-01

    Calculating the amount of energy released by a fission reaction (fission Q value) and the heating rate distribution in a nuclear reactor is an important part of the safety analysis. However, these calculations can become very complex. One of the codes that can be used for this type of analyses is the Monte Carlo transport code MCNP5. Currently it is impossible to calculate the Q value and heating rate disposition for delayed beta and delayed gamma particles directly from MCNP5. The purpose of this paper is to outline a rigorous method for indirectly calculating the Q values and heating rates in the High Flux Isotope Reactor (HFIR), based on previous similar studies carried out for very high-temperature reactor configurations. This method has been applied in this study to calculate heating rates for the beginning of cycle (BOC) and end-of-cycle (EOC) states of HFIR. In addition, the BOC results obtained for HFIR are compared with corresponding results for the Advanced Test Reactor. The fission Q value for HFIR was calculated as 200.2 MeV for the BOC and 201.3 MeV for the EOC. It was also determined that 95.1% and 95.4% of the heat was deposited within the HFIR fuel plates for the BOC and EOC models, respectively. This methodology can also be used for heating rate calculations for HFIR experiments. (authors)

  10. 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.

  11. 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

  12. Transmutation of MA in the high flux thermal reactor

    NASA Astrophysics Data System (ADS)

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

    2013-06-01

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

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

    NASA Astrophysics Data System (ADS)

    Iracane, Daniel; Chaix, Pascal; Alamo, Ana

    2008-04-01

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

  14. 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.

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

    SciTech Connect

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

    2011-07-01

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

  16. 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.

  17. 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.

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

    SciTech Connect

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

    2011-05-27

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

  19. 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)

  20. 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.

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

    SciTech Connect

    Sozer, A.

    1989-08-01

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

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

    SciTech Connect

    Rothrock, R.B.

    1990-06-01

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

  3. 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.

  4. Very high flux research reactors based on particle fuels

    SciTech Connect

    Powell, J.R.; Takahashi, H.

    1985-01-01

    A new approach to high flux research reactors is described, the VHFR (Very High Flux Reactor). The VHFR fuel region(s) are packed beds of HTGR-type fuel particles through which coolant (e.g., D/sub 2/O) flows directly. The small particle diameter (typically on the order of 500 microns) results in very large surface areas for heat transfer (approx. 100 cm/sup 2//cm/sup 3/ of bed), high power densities (approx. 10 megawatts per liter), and minimal ..delta..T between fuel and coolant (approx. 10 K) VHFR designs are presented which achieve steady-state fluxes of approx. 2x10/sup 16/ n/cm/sup 2/sec. Deuterium/beryllium combinations give the highest flux levels. Critical mass is low, approx. 2 kg /sup 235/U for 20% enriched fuel. Refueling can be carried out continuously on-line, or in a batch process with a short daily shutdown. Fission product inventory is very low, approx. 100 to 300 grams, depending on design.

  5. 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.

  6. Conversion feasibility studies for the Grenoble high flux reactor

    SciTech Connect

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

    1989-01-01

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

  7. Ultra-high flux reactor design using plate fuel technology

    SciTech Connect

    Lake, J.A.; Parsons, D.K.; Ryskamp, J.M.; Liebenthal, J.L.; Fillmore, G.N.

    1986-01-01

    The need for a new steady-state thermal neutron source of unprecedented intensity for materials science, isotope production, and fundamental physics research has been the subject of numerous national meetings and discussions. The challenge put forth by the research community is to produce a thermal neutron flux of 10/sup 16/ n/cm/sup 2/ x s in a large accessible volume with minimum fast neutron and gamma contamination. Ultra-high-flux reactor designs based on well-characterized plate fuel technologies have been examined. A double donut core configuration extends the range of peak operating conditions, which are traditionally limited by fuel plate temperatures and thermal-hydraulic conditions in the hot channel, to a point where these flux intensity goals can be attained.

  8. 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.

  9. Parameter trade-off studies for the ultra-high flux double donut reactor

    SciTech Connect

    Ryskamp, J.M.; Oh, C.H.; Ambrosek, R.G.; Lussie, W.G.; Lake, J.A.; Wadkins, R.P.

    1987-01-01

    The Idaho National Engineering Laboratory (INEL) is developing a reactor design in a double donut configuration for the advanced neutron source (ANS). The ANS, with a neutron flux ten times that of existing sources, is needed for materials science, isotope production, and fundamental physics research. The purpose of this study is to examine the reactor parameters for the INEL Ultrahigh Flux Reactor (UHFR) and recommend ways to optimize the reactor design.

  10. 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.

  11. 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…

  12. 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…

  13. 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.

  14. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2009-07-01

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

  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. Test calculations of the neutron flux on VVER-1000 reactor pressure vessel

    SciTech Connect

    Ilieva, K.D.; Belousov, S.I.; Antonov, S.Y.; Zaritsky, S.M.; Brodkin, E.B.

    1994-12-31

    A three dimensional test for calculation of the neutron fluence onto the VVER-1000 reactor pressure vessel (RPV) is presented. The test is based on the commercial VVER-1000 reactor design data. The flux results obtained by different authors are in good agreement.

  18. Reactor Engineering

    NASA Astrophysics Data System (ADS)

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

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

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

    SciTech Connect

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

    1998-06-01

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

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

    SciTech Connect

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

    1987-09-01

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

  1. 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.

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

    SciTech Connect

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

    2000-12-31

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

  3. 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

  4. 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}.

  5. 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.

  6. 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

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

    DOE PAGES

    Bess, John D.; Fujimoto, Nozomu

    2014-10-09

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

  8. 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.

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

    SciTech Connect

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

    2002-02-26

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

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

    SciTech Connect

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

    1987-01-01

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

  11. 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.

  12. 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

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

    SciTech Connect

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

    2016-10-01

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

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

    SciTech Connect

    Not Available

    1982-09-01

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

  15. Recent results of high heat flux testing at the Plasma Materials Test Facility

    NASA Astrophysics Data System (ADS)

    Youchison, Dennis L.; Watson, Robert D.; Marshall, Theron D.; McDonald, Jimmie M.

    1996-11-01

    High heat flux testing for the US fusion power program is the primary mission of the Plasma Materials Test Facility (PMTF) located at Sandia National Laboratory. This facility, an official Department of Energy User Facility, has been in operation for over 15 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 at Princeton, the DIII-D tokamak at General Atomics, and Alcator C-Mod at MIT, 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 testes, braze reliability tests, and safety related tests. The program's main focus now is on testing of beryllium and tungsten armor tiles bonded to divertor, limiter, and first wall components for the International Thermonuclear Experimental Reactor (ITER). The ITER project is a collaboration among the US, EU, RF, and Japanese fusion programs. This article provides a brief overview of the high heat flux testing capabilities at the PMTF, and describes some recent test results.

  16. Recent results of high heat flux testing at the Plasma Materials Test Facility

    SciTech Connect

    Youchison, D.L.; Watson, R.D.; Marshall, T.D.; McDonald, J.M.

    1996-12-31

    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 in Albuquerque, New Mexico. This facility, an official Department of Energy User Facility, has been in operation for over 15 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, the DIII-D tokamak pt General Atomics, and Alcator C-Mod at MIT, 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 program`s main focus now is on testing of beryllium and tungsten armor tiles bonded to divertor, limiter, and first wall components for the International Thermonuclear Experimental Reactor (ITER). The ITER project is a collaboration among the US, EU, RF, and Japanese fusion programs. This article provides a brief overview of the high heat flux testing capabilities at the PMTF, and describes some recent test results.

  17. 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.

  18. Dissolution Flowsheet for High Flux Isotope Reactor Fuel

    SciTech Connect

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

    2016-09-27

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

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

    NASA Astrophysics Data System (ADS)

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

    1990-04-01

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

  20. 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

  1. 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.

  2. 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.

  3. The method of life extension for the High Flux Isotope Reactor vessel

    NASA Astrophysics Data System (ADS)

    Chang, Shib-Jung

    1995-02-01

    The state of the vessel steel embrittlement as a result of neutron irradiation can be measured by its increase in the nil ductility temperature (NDT). This temperature is sometimes referred to as the brittle-ductile transition temperature (DBT) for fracture. The life extension of the High Flux Isotope Reactor (HFIR) vessel is calculated by using the method of fracture mechanics. A hydrostatic pressure test (hydrotest) is performed in order to determine a safe vessel static pressure. It is then followed by using fracture mechanics to project the reactor life from the safe hydrostatic pressure. The life extension calculation provides the following information on the remaining life of the reactor as a function of the nil ductility temperature increase: the probability of vessel fracture due to hydrotest vs vessel life at several hydrotest pressures; the hydrotest time interval vs the uncertainty of the nil ductility temperature increase rate; and the hydrotest pressure vs the uncertainty of the nil ductility temperature increase rate. It is understood that the use of a complete range of uncertainties of the nil ductility temperature increase is equivalent to the entire range of radiation damage that can be experienced by the vessel steel. From the numerical values for the probabilities of the vessel fracture as a result of hydrotest, it is estimated that the reactor vessel life can be extended up to 50 EFPY (100 MW) with the minimum vessel operating temperature equal to 85 F.

  4. 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.

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

    SciTech Connect

    Coyne, D.W.

    2008-07-01

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

  6. 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.

  7. 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.

  8. Specifications for high flux isotope reactor fuel elements HFIR-FE-3

    SciTech Connect

    Bowden, G.A.; Knight, R.W.

    1984-08-01

    This specification covers requirements for two types of aluminum-base fuel elements which together will be used as the fuel assembly in the High Flux Isotope Reactor (HFIR). Requirements are included for materials of construction, fabrication, assembly, inspection, and quality control to produce fuel elements in accordance with Company drawings.

  9. Probabilistic fracture mechanics analysis for the life extension estimate of the high flux isotope reactor vessel

    SciTech Connect

    Chang, S.J.

    1997-05-01

    The state of the vessel steel embrittlement as a result of neutron irradiation can be measured by its increase in the nil ductility temperature (NDT). This temperature is sometimes referred to as the brittle-ductile transition temperature (DBT) for fracture. The life extension of the High Flux Isotope Reactor (HFIR) vessel is calculated by using the method of fracture mechanics. A new method of fracture probability calculation is presented in this paper. The fracture probability as a result of the hydrostatic pressure test (hydrotest) is used to determine the life of the vessel. The hydrotest is performed in order to determine a safe vessel static pressure. It is then followed by using fracture mechanics to project the safe reactor operation time from the time of the satisfactory hydrostatic test. The life extension calculation provides the following information on the remaining life of the reactor as a function of the NDT increase: (1) the life of the vessel is determined by the probability of vessel fracture as a result of hydrotest at several hydrotest pressures and vessel embrittlement conditions, (2) the hydrotest time interval vs the NDT increase rate, and (3) the hydrotest pressure vs the NDT increase rate. It is understood that the use of a complete range of uncertainties of the NDT increase is equivalent to the entire range of radiation damage that can be experienced by the vessel steel. From the numerical values for the probabilities of the vessel fracture as a result of hydrotest, it is estimated that the reactor vessel life can be extended up to 50 EFPY (100 MW) with the minimum vessel operating temperature equal to 85{degrees}F.

  10. FASTER Test Reactor Preconceptual Design Report

    SciTech Connect

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

    2016-03-31

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

  11. 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.

  12. A neutronic feasibility study for LEU conversion of the high flux beam reactor (HFBR).

    SciTech Connect

    Pond, R. B.

    1998-01-16

    A neutronic feasibility study for converting the High Flux Beam Reactor at Brookhaven National Laboratory from HEU to LEU fuel was performed at Argonne National Laboratory. The purpose of this study is to determine what LEU fuel density would be needed to provide fuel lifetime and neutron flux performance similar to the current HEU fuel. The results indicate that it is not possible to convert the HFBR to LEU fuel with the current reactor core configuration. To use LEU fuel, either the core needs to be reconfigured to increase the neutron thermalization or a new LEU reactor design needs to be considered. This paper presents results of reactor calculations for a reference 28-assembly HEU-fuel core configuration and for an alternative 18-assembly LEU-fuel core configuration with increased neutron thermalization. Neutronic studies show that similar in-core and ex-core neutron fluxes, and fuel cycle length can be achieved using high-density LEU fuel with about 6.1 gU/cm{sup 3} in an altered reactor core configuration. However, hydraulic and safety analyses of the altered HFBR core configuration needs to be performed in order to establish the feasibility of this concept.

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

    SciTech Connect

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

    2016-12-21

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

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

    DOE PAGES

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

    2016-12-21

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

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

    SciTech Connect

    Chang, S.J.

    1998-08-01

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

  16. 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.

  17. 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.

  18. Neutron spectra at different High Flux Isotope Reactor (HFIR) pressure vessel surveillance locations

    SciTech Connect

    Remec, I.; Kam, F.B.

    1993-12-01

    This project addresses the potential problem of radiation embrittlement of reactor pressure vessel (RPV) supports. Surveillance specimens irradiated at the High Flux Isotope Reactor (HFIR) at relatively low neutron flux levels (about 1.5E + 8 cm{sup {minus}2}.s{sup {minus}1}) and low temperatures (about 50{degrees}C) showed embrittlement more rapidly than expected. Commercial power reactors have similar flux levels and temperatures at the level vessel support structures. The purposes of this work are to provide the neutron fluence spectra data that are needed to evaluate previously measured mechanical property changes in the HFIR, to explain the discrepancies in neutron flux levels between the nickel dosimeters and two other dosimeters, neptunium and beryllium, and to address any questions or peculiarities of the HFIR reactor environment. The current work consists of neutron and gamma transport calculations, dosimetry measurements, and least-squares logarithmic adjustment to obtain the best estimates for the neutron spectra and the related neutron exposure parameters. The results indicate that the fission rates in neptunium-237 (Np-237) and uranium-238 (U-238) and the helium production rates in beryllium-9 (Be-9) are dominated by photo-induced reactions. The displacements per atom rate for iron (dpa/s) from gamma rays is five times higher than the dpa/s from neutrons. The neutron fluxes in key 7, position 5 do not show any significant gradient in the surveillance capsule, but key 4 and key 2 showed differences in magnitude as well as in the shape of the spectrum. The stainless steel monitor in the V-notch of the Charpy specimens of the surveillance capsules is adequate to determine the neutron flux above 1.0 MeV at the desired V-notch location. Simultaneous adjustment of neutron and gamma fluxes with the measurements has been demonstrated and should avoid future problems with photo-induced reactions.

  19. 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.

  20. 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

  1. FASTER test reactor preconceptual design report summary

    SciTech Connect

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

    2016-02-29

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

  2. 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.

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

    NASA Astrophysics Data System (ADS)

    Watson, Robert D.

    1993-02-01

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

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

    SciTech Connect

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

    2016-04-01

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

  5. 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.

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

    NASA Astrophysics Data System (ADS)

    Pal, Usha; Jagannathan, V.

    2007-02-01

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

  7. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2010-11-01

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

  9. Application of the successive linear programming technique to the optimum design of a high flux reactor using LEU fuel

    SciTech Connect

    Mo, S.C.

    1991-01-01

    The successive linear programming technique is applied to obtain the optimum thermal flux in the reflector region of a high flux reactor using LEU fuel. The design variables are the reactor power, core radius and coolant channel thickness. The constraints are the cycle length, average heat flux and peak/average power density ratio. The characteristics of the optimum solutions with various constraints are discussed.

  10. Application of the successive linear programming technique to the optimum design of a high flux reactor using LEU fuel

    SciTech Connect

    Mo, S.C.

    1991-12-31

    The successive linear programming technique is applied to obtain the optimum thermal flux in the reflector region of a high flux reactor using LEU fuel. The design variables are the reactor power, core radius and coolant channel thickness. The constraints are the cycle length, average heat flux and peak/average power density ratio. The characteristics of the optimum solutions with various constraints are discussed.

  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. 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.

  13. 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

  14. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-09-01

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

  16. 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.

  17. 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.

  18. 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.

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

    SciTech Connect

    Primm, R.T. III.

    1992-03-01

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

  20. 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.

  1. 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.

  2. 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.

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

    SciTech Connect

    Hassanein, A.M.

    1988-07-01

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

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

    SciTech Connect

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

    1994-12-31

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

  5. 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.

  6. 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.

  7. 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.

  8. 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.

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

    SciTech Connect

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

    1994-12-31

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

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

    SciTech Connect

    Field, Kevin G.; Howard, Richard H.

    2016-08-19

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

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

    SciTech Connect

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

    1986-08-01

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

  12. 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.; Ye, M. Y.; 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.

  13. 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.

  14. 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.

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

    SciTech Connect

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

    1981-10-01

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

  16. 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.

  17. Fuel elements of research reactors in China

    SciTech Connect

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

    1988-01-01

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

  18. Engineering and erection of a 300kW high-flux solar simulator

    NASA Astrophysics Data System (ADS)

    Wieghardt, Kai; Laaber, Dmitrij; Hilger, Patrick; Dohmen, Volkmar; Funken, Karl-Heinz; Hoffschmidt, Bernhard

    2017-06-01

    German Aerospace Center (DLR) is currently constructing a new high-flux solar simulator synlight which shall be commissioned in late 2016. The new facility will provide three separately operated experimental spaces with expected radiant powers of about 300kW / 240kW / 240kW respectively. synlight was presented to the public for the first time at SolarPACES 2015 [1]. Its engineering and erection is running according to plan. The current presentation reports about the engineering and the ongoing erection of the novel facility, and gives an outlook on its new level of possibilities for solar testing and qualification.

  19. 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.

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

    SciTech Connect

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

    2016-09-01

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

  1. 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.

  2. 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).

  3. 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.

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

    SciTech Connect

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

    1994-12-31

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

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

    SciTech Connect

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

    1985-10-01

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

  6. Progress towards boron neutron capture therapy at the High Flux Reactor Petten.

    PubMed

    Moss, R L

    1990-01-01

    During 1988 the first positive steps were taken to proceed with the design and construction of a neutron capture therapy facility on the High Flux Reactor (HFR) at Petten. The immediate aim is to realise within a short time (summer 1989), an epithermal neutron beam for radiobiological and filter optimisation studies on one of the 10 small aperture horizontal beam tubes. The following summer, a much larger neutron beam, i.e., in cross section and neutron fluence rate, will be constructed on one of the two large beam tubes that replaced the old thermal column in 1984. This latter beam tube faces one whole side of the reactor vessel, extending from a 50 x 40 cm input aperture to a 35 x 35 cm exit hole. The radiotherapeutic facility will be housed here, with the intention to start clinical trials at the beginning of 1991. This paper describes the present status of the project and includes: a general description of the pertinent characteristics with respect to NCT of the HFR; results of the recently completed preliminary neutron metrology and computer modeling at the input end of the candidate beam tube; the structure and planning of the proposed Work Programme; and the respective direct and indirect participation and collaboration with the Netherlands Cancer Institute and the European Collaboration Group on BNCT.

  7. 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.

  8. 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

  9. 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

  10. 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.

  11. 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.

  12. 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.

  13. 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

  14. 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).

  15. 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.

  16. 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.

  17. 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)

  18. 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.

  19. 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

  20. Hydrogen Cylinder Storage Array Explosion Evaluations at the High Flux Isotope Reactor

    SciTech Connect

    Cook, David Howard; Griffin, Frederick P; Hyman III, Clifton R

    2010-01-01

    The safety analysis for a recently-installed cold neutron source at the High Flux Isotope Reactor (HFIR) involved evaluation of potential explosion consequences from accidental hydrogen jet releases that could occur from an array of hydrogen cylinders. The scope of the safety analysis involved determination of the release rate of hydrogen, the total quantity of hydrogen assumed to be involved in the explosion, the location of an ignition point or center of the explosion from receptors of interest, and the peak overpressure at the receptors. To evaluate the total quantity of hydrogen involved in the explosion, a 2D model was constructed of the jet concentration and a radial-axial integral over the jet cloud from the centerline to the flammability limit of 4% was used to determine the hydrogen mass to be used as a source term. The location of the point source was chosen as the peak of the jet centerline concentration profile. Consequences were assessed using a combination of three methods for estimating local overpressure as a function of explosion source strength and distance: the Baker-Strehlow method, the TNT-equivalence method, and the TNO method. Results from the explosions were assessed using damage estimates in screening tables for buildings and industrial equipment.

  1. 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.

  2. 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.

  3. 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.

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

    SciTech Connect

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

    2016-11-01

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

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

    NASA Astrophysics Data System (ADS)

    Araki, Masanori

    1993-03-01

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

  6. Deterministic Modeling of the High Temperature Test Reactor

    SciTech Connect

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

    2010-06-01

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

  7. 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)

  8. 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)

  9. 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).

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

    SciTech Connect

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

    2010-02-01

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

  11. 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.

  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. High-Fidelity Modelng and Simulation for a High Flux Isotope Reactor Low-Enriched Uranium Core Design

    DOE PAGES

    Betzler, Benjamin R.; Chandler, David; Davidson, Eva E.; ...

    2017-05-08

    A high-fidelity model of the High Flux Isotope Reactor (HFIR) with a low-enriched uranium (LEU) fuel design and a representative experiment loading has been developed to serve as a new reference model for LEU conversion studies. With the exception of the fuel elements, this HFIR LEU model is completely consistent with the current highly enriched uranium HFIR model. Results obtained with the new LEU model provide a baseline for analysis of alternate LEU fuel designs and further optimization studies. The newly developed HFIR LEU model has an explicit representation of the HFIR-specific involute fuel plate geometry, including the within-plate fuelmore » meat contouring, and a detailed geometry model of the fuel element side plates. Such high-fidelity models are necessary to accurately account for the self-shielding from 238U and the depletion of absorber materials present in the side plates. In addition, a method was developed to account for fuel swelling in the high-density LEU fuel plates during the depletion simulation. In conclusion, calculated time-dependent metrics for the HFIR LEU model include fission rate and cumulative fission density distributions, flux and reaction rates for relevant experiment locations, point kinetics data, and reactivity coefficients.« less

  14. 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].

  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. Mixed oxide fuel testing capabilities in the Advanced Test Reactor

    SciTech Connect

    Ryskamp, J.M.; Sterbentz, J.W.; Chang, G.S.; Terry, W.K.

    1996-08-01

    The most attractive way to dispose of weapons-grade Plutonium (WGPu) is to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel--i.e., plutonia (PuO{sub 2}) mixed with urania (UO{sub 2}). Before US reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) possesses many advantages for performing tests to resolve most of the issues. It has ample core test volume, high neutron flux, test loops with cooling systems independent of the core coolant, and extensive support facilities. The ATR can deliver a neutron flux of appropriate intensity and energy distribution to the MOX test specimens while simultaneously accommodating test requirements for other programs. The authors have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their design values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. Furthermore, this data can be obtained more quickly by using ATR instead of testing in a commercial LWR.

  17. 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

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

    SciTech Connect

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

    2016-08-01

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

  19. 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.

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

    SciTech Connect

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

    2016-03-01

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

  1. Ultrahigh flux double donut research reactor design

    SciTech Connect

    Ryskamp, J.M.; Parsons, D.K.; Lake, J.A.

    1986-01-01

    A new steady-state thermal neutron source of unprecedented intensity is under development for materials science research, isotope production, and fundamental physics research. The challenge put forth by the research community is to produce a thermal neutron flux of 10/sup 16/ n/(cm/sup 2/s) in a large accessible volume with minimum fast neutron and gamma contamination. Ultrahigh flux reactor designs based on well-characterized plate-fuel technologies have been examined at the Idaho National Engineering Laboratory. A ''double donut'' core configuration extends the range of peak operating conditions, which are traditionally limited by fuel plate temperatures and thermal hydraulic conditions in the hot channel, to a point where these flux intensity goals can be attained.

  2. 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.

  3. 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

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

    SciTech Connect

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

    2016-11-01

    Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations. Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.

  5. High energy reactor neutrinos

    NASA Astrophysics Data System (ADS)

    Raper, Neill

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

  6. 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

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

    SciTech Connect

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

    1983-12-02

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

  9. Report detailing comparative analysis of results from high flux isotope reactor and national institute of standards technology small-angle neutron scattering experiments

    SciTech Connect

    Sokolov, Mikhail A.; Littrell, Ken; Wells, Peter; Cunningham, Nicholas J.

    2015-09-01

    The major issues regarding irradiation effects are discussed in [1-3] and have also been discussed in previous progress and milestone reports. As noted previously, of the many significant issues discussed, the issue considered to have the most impact on the current regulatory process is that associated with effects of neutron irradiation on RPV steels at high fluence, for long irradiation times, and as affected by neutron flux. It is clear that embrittlement of RPV steels is a critical issue that may limit LWR plant life extension. The primary objective of the LWRSP RPV task is to develop robust predictions of transition temperature shifts (TTS) at high fluence ( t) to at least 1020 n/cm2 (>1 MeV) pertinent to plant operation of some pressurized water reactors (PWR) for 80 full power years. Correlations between the high flux test reactor results and low flux surveillance specimens must be established for proper RPV embrittlement predictions of the current nuclear power fleet. Additionally, a complete understanding of defect evolution for high nickel RPV steels is needed to characterize the embrittlement potential of Mn-Ni-enriched precipitates (MNPs), particularly for the high fluence regime. While understanding of copper-enriched precipitates (CRPs) have been fully developed, the recent discovery and experimental verification [4] of late blooming MNPs with little to no copper for nucleation has stimulated research efforts to understand the evolution of these phases. New and existing databases will be combined to support developing physically based models of TTS for high fluence-low flux ( < 10 11n/cm2-s) conditions, beyond the existing surveillance database, to neutron fluences of at least 1 1020 n/cm2 (>1 MeV). Moreover, large number of various RPV materials have been irradiated in ATR-2 experiment and will be jointly studied by University of California Santa Barbara (UCSB) and ORNL to address majority of microstructural characteristics

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. Heat Flux Sensor Testing

    NASA Astrophysics Data System (ADS)

    Clark, D. W.

    2002-07-01

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

  16. 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.

  17. 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%).

  18. 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.

  19. 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.

  20. 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.

  1. 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)

  2. 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.

  3. Calculation of neutron and gamma fluxes in support to the interpretation of measuring devices irradiated in the core periphery of the OSIRIS Material Testing Reactor

    SciTech Connect

    Malouch, Fadhel

    2015-07-01

    Technological irradiations carried out in material testing reactors (MTRs) are used to study the behavior of materials under irradiation conditions required by different types of nuclear power plants (NPPs). For MTRs, specific instrumentation is required for the experiment monitoring and for the characterization of irradiation conditions, in particular the flux of neutrons and photons. To measure neutron and photon flux in experimental locations, different sensors can be used, such as SPNDs (self-powered neutron detectors), SPGDs (self-powered gamma detectors) and ionization chambers. These sensors involve interactions producing ultimately a measurable electric current. Various sensors have been recently tested in the core periphery of the OSIRIS reactor (located at the CEA-Saclay center) in order to qualify their responses to the neutron and the photon flux. One of the key input data for this qualification is to have a relevant evaluation of neutron and gamma fluxes at the irradiation location. The objective of this work is to evaluate the neutron and the gamma flux in the core periphery of the OSIRIS reactor. With this intention, specific neutron-photonic three-dimensional calculations have been performed and are mainly based on the TRIPOLI-4{sup R} three-dimensional continuous-energy Monte Carlo code, developed by CEA (Saclay Center) and extensively validated against reactor dosimetry benchmarks. In the case of the OSIRIS reactor, TRIPOLI-4{sup R} code has been validated against experimental results based on neutron flux and nuclear heating measurements performed in ex-core and in-core experiments. In this work, simultaneous contribution of neutrons and gamma photons in the core periphery is considered using neutron-photon coupled transport calculations. Contributions of prompt and decay photons have been taken into account for the gamma flux calculation. Specific depletion codes are used upstream to provide the decay-gamma sources required by TRIPOLI-4

  4. 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.

  5. 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.

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

    SciTech Connect

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

    2014-07-01

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

  7. 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.

  8. Advanced Test Reactor Tour

    ScienceCinema

    Miley, Don

    2016-07-12

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

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

    SciTech Connect

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

    2016-11-01

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

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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

  16. High-Speed Tests of Radial-Engine Cowlings

    NASA Technical Reports Server (NTRS)

    Robinson, Russell G.; Becker, John V.

    1939-01-01

    The drag characteristics of eight radial-engine cowlings have been determined over a wide speed range in the N.A.C.A. 8-foot high-speed wind tunnel. The pressure distribution over all cowlings was measured, to and above the speed of the compressibility burble, as an aid in interpreting the force tests. One-fifth-scale models of radial-engine cowlings on a wing-nacelle combination mere used in the tests.

  17. High-Speed Tests of Conventional Radial-Engine Cowlings

    NASA Technical Reports Server (NTRS)

    Robinson, Russell G; Becker, John V

    1942-01-01

    The drag characteristics of eight radial-engine cowlings have been determined over a wide speed range in the NACA 8-foot high-speed wind tunnel. The pressure distribution over all cowlings was measured, to and above the speed of the compressibility burble, as an aid in interpreting the force tests. One-fifth-scale models of radial-engine cowlings on a wing-nacelle combination were used in the tests.

  18. 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.

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

    NASA Technical Reports Server (NTRS)

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

    1992-01-01

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

  20. Dissecting Reactor Antineutrino Flux Calculations.

    PubMed

    Sonzogni, A A; McCutchan, E A; Hayes, A C

    2017-09-15

    Current predictions for the antineutrino yield and spectra from a nuclear reactor rely on the experimental electron spectra from ^{235}U, ^{239}Pu, ^{241}Pu and a numerical method to convert these aggregate electron spectra into their corresponding antineutrino ones. In the present work we investigate quantitatively some of the basic assumptions and approximations used in the conversion method, studying first the compatibility between two recent approaches for calculating electron and antineutrino spectra. We then explore different possibilities for the disagreement between the measured Daya Bay and the Huber-Mueller antineutrino spectra, including the ^{238}U contribution as well as the effective charge and the allowed shape assumption used in the conversion method. We observe that including a shape correction of about +6%  MeV^{-1} in conversion calculations can better describe the Daya Bay spectrum. Because of a lack of experimental data, this correction cannot be ruled out, concluding that in order to confirm the existence of the reactor neutrino anomaly, or even quantify it, precisely measured electron spectra for about 50 relevant fission products are needed. With the advent of new rare ion facilities, the measurement of shape factors for these nuclides, for many of which precise beta intensity data from TAGS experiments already exist, would be highly desirable.

  1. Dissecting Reactor Antineutrino Flux Calculations

    NASA Astrophysics Data System (ADS)

    Sonzogni, A. A.; McCutchan, E. A.; Hayes, A. C.

    2017-09-01

    Current predictions for the antineutrino yield and spectra from a nuclear reactor rely on the experimental electron spectra from 235U, 239Pu, 241Pu and a numerical method to convert these aggregate electron spectra into their corresponding antineutrino ones. In the present work we investigate quantitatively some of the basic assumptions and approximations used in the conversion method, studying first the compatibility between two recent approaches for calculating electron and antineutrino spectra. We then explore different possibilities for the disagreement between the measured Daya Bay and the Huber-Mueller antineutrino spectra, including the 238U contribution as well as the effective charge and the allowed shape assumption used in the conversion method. We observe that including a shape correction of about +6 % MeV-1 in conversion calculations can better describe the Daya Bay spectrum. Because of a lack of experimental data, this correction cannot be ruled out, concluding that in order to confirm the existence of the reactor neutrino anomaly, or even quantify it, precisely measured electron spectra for about 50 relevant fission products are needed. With the advent of new rare ion facilities, the measurement of shape factors for these nuclides, for many of which precise beta intensity data from TAGS experiments already exist, would be highly desirable.

  2. Dissecting Reactor Antineutrino Flux Calculations

    DOE PAGES

    Sonzogni, A. A.; McCutchan, E. A.; Hayes, A. C.

    2017-09-15

    Current predictions for the antineutrino yield and spectra from a nuclear reactor rely on the experimental electron spectra from 235 U , 239 Pu , 241 Pu and a numerical method to convert these aggregate electron spectra into their corresponding antineutrino ones. In our present work we investigate quantitatively some of the basic assumptions and approximations used in the conversion method, studying first the compatibility between two recent approaches for calculating electron and antineutrino spectra. We then explore different possibilities for the disagreement between the measured Daya Bay and the Huber-Mueller antineutrino spectra, including the 238 U contribution as wellmore » as the effective charge and the allowed shape assumption used in the conversion method. Here, we observe that including a shape correction of about + 6 % MeV - 1 in conversion calculations can better describe the Daya Bay spectrum. Because of a lack of experimental data, this correction cannot be ruled out, concluding that in order to confirm the existence of the reactor neutrino anomaly, or even quantify it, precisely measured electron spectra for about 50 relevant fission products are needed. With the advent of new rare ion facilities, the measurement of shape factors for these nuclides, for many of which precise beta intensity data from TAGS experiments already exist, would be highly desirable.« less

  3. 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

  4. 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

  5. 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

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

    SciTech Connect

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

    1999-12-01

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

  7. 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

  8. 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.

  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. 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.

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

    SciTech Connect

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

    1982-01-01

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

  12. 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.

  13. 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

  14. 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.

  15. 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.

  16. Evaluation of cooling concepts and specimen geometries for high heat flux tests on neutron irradiated divertor elements

    SciTech Connect

    Linke, J.; Bolt. H.; Breitbach, G.

    1994-12-31

    To assess the lifetime and the long term heat removal capabilities of plasma facing components in future thermonuclear fusion reactors such as ITER, neutron irradiation and subsequent high heat flux tests will be most essential. The effect of neutron damage will be simulated in material test reactors (such as the HFR-Petten) in a fission neutron environment. To investigate the heat loads during normal and off-normal operation scenarios a 60 kW electron beam test stand (Juelich Divertor Test Facility in Hot Cells, JUDITH) has been installed in a hot cell which can be operated by remote handling techniques. In this facility inertially cooled test coupons can be handled as well as small actively cooled divertor mock-ups. A special clamping mechanism for small test coupons (25 mm x 25 mm x 35 mm) with an integrated coolant channel within a copper or TZM heat sink has been developed and tested in an electron beam test bed. This method is an attractive alternative to costly large scale tests on complete divertor modules. The temperature and stress fields in individual CFC or beryllium tiles brazed to metallic heat sink (e.g. copper or TZM) can be investigated before and after neutron irradiation with moderate efforts.

  17. 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

  18. Parameter trade-off studies for the ultrahigh flux double donut reactor

    SciTech Connect

    Ryskamp, J.M.; Oh, C.H.; Ambrosek, R.G.; Lussie, W.G.; Lake, J.A.; Wadkins, R.P.

    1987-01-01

    The Idaho National Engineering Laboratory (INEL) is developing a reactor design in a double donut configuration for the advanced neutron source (ANS). The ANS, with a neutron flux ten times that of existing sources, is needed for materials science, isotope production, and fundamental physics research. The purpose of this study is to examine the reactor parameters for the INEL Ultrahigh Flux Reactor (UHFR) and recommend ways to optimize the reactor design.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

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

    SciTech Connect

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

    2000-07-18

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

  8. 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.

  9. Considerations in the development of a process to manufacture low-enriched uranium foil fuel for the high flux isotope reactor

    SciTech Connect

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

    2008-07-15

    A reference flow sheet is the one of the first planning steps in the development of a manufacturing capacity for low enriched uranium foil fuels and can be used to develop a work structure, a critical path schedule and identify development needs. The reference flow sheet presented is specific to the High Flux Isotope Reactor and is used to estimate the change in HFIR operating cost due to fuel conversion. (author)

  10. 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.

  11. 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.

  12. Reactor antineutrino fluxes – Status and challenges

    DOE PAGES

    Huber, Patrick

    2016-04-22

    Here, 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.

  13. 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).

  14. 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.

  15. 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.

  16. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

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

  18. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-10-01

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

  20. 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.

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

    NASA Astrophysics Data System (ADS)

    Budaev, V. P.

    2016-12-01

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

  2. 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.

  3. 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

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

    NASA Technical Reports Server (NTRS)

    Noack, W G

    1920-01-01

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

  5. 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

  6. 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

  7. 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.

  8. 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.

  9. Designing a Gas Test Loop for the Advanced Test Reactor

    SciTech Connect

    James R. Parry

    2005-11-01

    The Generation IV Reactor Program and the Advanced Fuel Cycle Initiative are investigating some new reactor concepts which require extensive materials and fuels testing in a fast neutron spectrum. The capability to test materials and fuels in a fast neutron flux in the United States is very limited to non-existent. It has been proposed to install a gas test loop (GTL) in one of the lobes of the Advanced Test Reactor (ATR) at the Idaho National Laboratory and harden the spectrum to provide some fast neutron flux testing capabilities in the United States. This paper describes the neutronics investigation into the design of the GTL for the ATR.

  10. 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

  11. Engineering test facility

    SciTech Connect

    Steiner, D.; Becraft, W.R.; Sager, P.H.

    1981-01-01

    The vehicle by which the fusion program would move into the engineering testing phase of fusion power development is designated the Engineering Test Facility (ETF). The ETF would provide a test-bed for reactor components in the fusion environment. In order to initiate preliminary planning for the ETF decision, the Office of Fusion Energy established the ETF Design Center activity to prepare the design of the ETF. This paper described the design status of the ETF.

  12. Determining reactor flux from xenon-136 and cesium-135 in spent fuel

    NASA Astrophysics Data System (ADS)

    Hayes, A. C.; Jungman, Gerard

    2012-10-01

    The ability to infer reactor flux from spent fuel or seized fissile material would enhance the tools of nuclear forensics and nuclear nonproliferation significantly. We show that reactor flux can be inferred from the ratios of xenon-136 to xenon-134 and cesium-135 to cesium-137. If the average flux of a reactor is known, the flux inferred from measurements of spent fuel could help determine whether that spent fuel was loaded as a blanket or close to the mid-plane of the reactor. The cesium ratio also provides information on reactor shutdowns during the irradiation of fuel, which could prove valuable for identifying the reactor in question through comparisons with satellite reactor heat monitoring data. We derive analytic expressions for these correlations and compare them to experimental data and to detailed reactor burn simulations. The enrichment of the original uranium fuel affects the correlations by up to 3%, but only at high flux.

  13. 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.

  14. 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

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

    NASA Technical Reports Server (NTRS)

    Timko, L. P.

    1984-01-01

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

  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. 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.

  18. 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.

  19. High temperature reactors

    NASA Astrophysics Data System (ADS)

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

    2008-12-01

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

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

    SciTech Connect

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

    2015-11-01

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

  1. 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.

  2. 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.

  3. Broad Application Test Reactor cost reduction study

    SciTech Connect

    Sekot, J.P.; Amidei, W.; Kologi, A.L.; Lopez, D.A.; McDaniel, J.H.; Mousseau, D.R.; Motloch, C.G.; Pierce, R.L.; Rogers, N.K.; Ryskamp, J.M.; Sherick, D.M.; Sorman, K.L.; Stanley, V.R.; Werner, J.E.

    1993-03-01

    A Broad Application Test Reactor (BATR) is needed to provide nuclear testing capacity and capability for future testing needs. The BATR concept is expected to consider a broad range of test applications. The present day costs of design and construction of new nuclear facilities demands that methods, processes, and requirements be critically reviewed and improved if these facilities are to be affordable in the future. The cost of major new projects in this country has been escalating faster than the rate of inflation. For example, the Advanced Test Reactor (ATR), a reactor built at the Idaho National Engineering Laboratory (INEL) in the early sixties, cost about $38M in 1960 dollars. Estimates to build that same reactor today range as high as $1B. Although somewhat different in mission, but in many respects comparable to theATR in function, size, and complexity, the Advanced Neutron Source (ANS) proposed by the Oak Ridge National Laboratory (ORNL), is currently estimated to cost about $2B for design and construction. This report presents cost concerns, cost reduction ideas, and recommendations that were identified and developed by the Cost Reduction Study Team for consideration by the BATR Design Team.

  4. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2006-11-01

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

  6. Internal Combustion Engines as Fluidized Bed Reactors

    NASA Astrophysics Data System (ADS)

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

    2016-11-01

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

  7. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2013-06-01

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

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

    NASA Technical Reports Server (NTRS)

    Krech, Robert H.; Caledonia, George E.

    1986-01-01

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

  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. FED reactor engineering features

    SciTech Connect

    Sager, P.H.; Brown, T.G.; Fuller, G.M.; Smith, G.E.

    1982-01-01

    The Fusion Engineering Device (FED) Baseline design incorporates a number of features which were selected to enhance its maintainability, as well as limit cost and achieve reliable operation. An installation of ten TF coils and ten torus sectors was selected on the basis of plasma chamber segmentation studies and TF coil cost tradeoff studies, permitting removal of a torus sector with a single radial motion. The design also features a shield sector support spool which provides a plasma chamber vacuum boundary and access to the shield sectors. The vacuum seals are made at the outboard face of the torus so that they can be readily cut and rewelded. A pumped limiter provides plasma edge definition and impurity control. Ten individual blades are inserted through the shield sector in an arrangement that permits replacement without sector removal. ICRH is used for plasma bulk heating. Two EF coils, which are located inside the TF coil bore, are segmented so that they can be removed if necessary. The removal of the superconducting lower outboard EF coil, which is trapped under the TF coil assembly, presents a problem; consideration is being given to increasing its diameter and relocating it so that it can be lifted up around the TF coils.

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

    SciTech Connect

    Chandler, David; Ellis, Ronald James

    2015-01-01

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

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

    SciTech Connect

    Harpenau, Evan M.

    2012-01-13

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

  14. 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.

  15. 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

  16. High flux heat exchanger

    NASA Astrophysics Data System (ADS)

    Flynn, Edward M.; Mackowski, Michael J.

    1993-01-01

    This interim report documents the results of the first two phases of a four-phase program to develop a high flux heat exchanger for cooling future high performance aircraft electronics. Phase 1 defines future needs for high flux heat removal in advanced military electronics systems. The results are sorted by broad application categories: (1) commercial digital systems, (2) military data processors, (3) power processors, and (4) radar and optical systems. For applications expected to be fielded in five to ten years, the outlook is for steady state flux levels of 30-50 W/sq cm for digital processors and several hundred W/sq cm for power control applications. In Phase 1, a trade study was conducted on emerging cooling technologies which could remove a steady state chip heat flux of 100 W/sq cm while holding chip junction temperature to 90 C. Constraints imposed on heat exchanger design, in order to reflect operation in a fighter aircraft environment, included a practical lower limit on coolant supply temperature, the preference for a nontoxic, nonflammable, and nonfreezing coolant, the need to minimize weight and volume, and operation in an accelerating environment. The trade study recommended the Compact High Intensity Cooler (CHIC) for design, fabrication, and test in the final two phases of this program.

  17. 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.

  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. Reactor Simulator Testing

    NASA Technical Reports Server (NTRS)

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

    2013-01-01

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

  20. 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.

  1. Operational safety at the fast flux test facility

    SciTech Connect

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

    1986-01-01

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

  2. Potential for new societal contributions from the advanced test reactor

    SciTech Connect

    Ryskamp, J.M.; Conner, J.E.; Ingram, F.W. )

    1993-01-01

    The mission of the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory is to study the effects of intense radiation on materials and fuels and to produce radioisotopes for the U.S. Department of Energy (DOE) for government and commercial applications. Because of reductions in defense spending, four of the nine loop test spaces will become available in 1994. The purpose of this paper is to explore the potential benefits to society from these available neutrons. The ATR is a 250-MW(thermal) light water reactor with highly enriched uranium in plate-type fuel. Forty fuel elements are arranged in a serpentine pattern. The ATR uses a combination of hafnium control drums and shim rods to adjust power and hold flux distortion to a minimum. The different quadrants of the ATR can be operated at significantly different power levels to meet a variety of mission requirements. Irradiation positions are available at various locations throughout the core and beryllium reflector.

  3. 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.

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

    SciTech Connect

    Budaev, V. P.

    2016-12-15

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

  5. Design, testing and modeling of a high-gain magnetic flux-compression generator

    SciTech Connect

    Sheppard, M.G.; Freeman, B.L.; Bowers, R.L.; Brownell, J.H.; Fowler, C.M.; Fritz, J.N.; Greene, A.E.; Marsh, S.P.; Oliphant, T.A.; Tubbs, D.L.

    1986-01-01

    Using a simultaneously initiated cylindrical explosive, a coaxial magnetic flux-compression generator (FCG) was designed to test high-current-gain limitations. A coaxial design with a lossless gain of approx.100:1 was chosen for its efficiency, relative simplicity, and calculability. Theoretical design included modeling as well as 1-D and 2-D hydrodynamic and MHD calculations. A 69.3-cm cylinder of PBX-9501 high explosive, 20.3 cm in diameter, was used to drive the Al armature into a Cu stator. The initial current supplied by a capacitor bank was approx.3 MA which produced a final current approx.75 MA. Details of the experiment and a comparison with calculations are presented.

  6. 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

  7. 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.

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

    NASA Astrophysics Data System (ADS)

    Boåsen, Magnus; Efsing, Pål; Ehrnstén, Ulla

    2017-02-01

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

  9. 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

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

    SciTech Connect

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

    2015-07-01

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

  11. Surface heat flux feedback controlled impurity seeding experiments with Alcator C-Mod’s high-Z vertical target plate divertor: performance, limitations and implications for fusion power reactors

    NASA Astrophysics Data System (ADS)

    Brunner, D.; Wolfe, S. M.; LaBombard, B.; Kuang, A. Q.; Lipschultz, B.; Reinke, M. L.; Hubbard, A.; Hughes, J.; Mumgaard, R. T.; Terry, J. L.; Umansky, M. V.; The Alcator C-Mod Team

    2017-08-01

    The Alcator C-Mod team has recently developed a feedback system to measure and control surface heat flux in real-time. The system uses real-time measurements of surface heat flux from surface thermocouples and a pulse-width modulated piezo valve to inject low-Z impurities (typically N2) into the private flux region. It has been used in C-Mod to mitigate peak surface heat fluxes  >40 MW m-2 down to  <10 MW m-2 while maintaining excellent core confinement, H 98  >  1. While the system works quite well under relatively steady conditions, use of it during transients has revealed important limitations on feedback control of impurity seeding in conventional vertical target plate divertors. In some cases, the system is unable to avoid plasma reattachment to the divertor plate or the formation of a confinement-damaging x-point MARFE. This is due to the small operational window for mitigated heat flux in the parameters of incident plasma heat flux, plasma density, and impurity density as well as the relatively slow response of the impurity gas injection system compared to plasma transients. Given the severe consequences for failure of such a system to operate reliably in a reactor, there is substantial risk that the conventional vertical target plate divertor will not provide an adequately controllable system in reactor-class devices. These considerations motivate the need to develop passively stable, highly compliant divertor configurations and experimental facilities that can test such possible solutions.

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

    NASA Technical Reports Server (NTRS)

    Nesbitt, J. A.

    1990-01-01

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

  13. 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.

  14. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2015-03-01

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

  16. 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.

  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. 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).

  19. Setup for polarized neutron imaging using in situ 3He cells at the Oak Ridge National Laboratory High Flux Isotope Reactor CG-1D beamline

    NASA Astrophysics Data System (ADS)

    Dhiman, I.; Ziesche, Ralf; Wang, Tianhao; Bilheux, Hassina; Santodonato, Lou; Tong, X.; Jiang, C. Y.; Manke, Ingo; Treimer, Wolfgang; Chatterji, Tapan; Kardjilov, Nikolay

    2017-09-01

    In the present study, we report a new setup for polarized neutron imaging at the ORNL High Flux Isotope Reactor CG-1D beamline using an in situ 3He polarizer and analyzer. This development is very important for extending the capabilities of the imaging instrument at ORNL providing a polarized beam with a large field-of-view, which can be further used in combination with optical devices like Wolter optics, focusing guides, or other lenses for the development of microscope arrangement. Such a setup can be of advantage for the existing and future imaging beamlines at the pulsed neutron sources. The first proof-of-concept experiment is performed to study the ferromagnetic phase transition in the Fe3Pt sample. We also demonstrate that the polychromatic neutron beam in combination with in situ 3He cells can be used as the initial step for the rapid measurement and qualitative analysis of radiographs.

  20. Design and performance of vacuum system for high heat flux test facility

    NASA Astrophysics Data System (ADS)

    Swamy Kidambi, Rajamannar; Mokaria, Prakash; Khirwadkar, Samir; Belsare, Sunil; Khan, M. S.; Patel, Tushar; Krishnan, Deepu S.

    2017-04-01

    High heat flux test facility (HHFTF) at IPR is used for testing thermal performance of plasma facing materials or components. It consists of various subsystems like vacuum system, high power electron beam system, diagnostic and calibration system, data acquisition and control system and high pressure high temperature water circulation system. Vacuum system consists of large D-shaped chamber, target handling system, pumping systems and support structure. The net volume of vacuum chamber is 5 m3 was maintained at the base pressure of the order of 10-6 mbar for operation of electron gun with minimum beam diameter which is achieved with turbo-molecular pump (TMP) and cryo pump. A variable conductance gate valve is used for maintaining required vacuum in the chamber. Initial pumping of the chamber was carried out by using suitable rotary and root pumps. PXI and PLC based faster real time data acquisition and control system is implemented for performing the various operations like remote operation, online vacuum data measurements, display and status indication of all vacuum equipments. This paper describes in detail the design and implementation of various vacuum system for HHFTF.

  1. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2016-08-01

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

  3. 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.

  4. 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

  5. Ultra high flux reactor design: probing the limits of plate fuel technology. [Thermal neutron flux exceeding 10/sup 16/ n/cm/sup 2/. s

    SciTech Connect

    Lake, J.A.; Parsons, D.K.; Liebenthal, J.L.; Ryskamp, J.M.; Fillmore, G.N.; deBoisblanc, D.R.

    1985-12-01

    This paper examined the capabilities and limitations of the plate fuel technology for producing unprecedented thermal neutron flux levels exceeding 10/sup 16/ n/cm/sup 2/.s. The requirements for a flux configured over a large volume, and accessible to experimental instruments, as well as isotope production and materials irradiation access, dictates that the flux peak be generated external to the core and, most efficiently, in a large D/sub 2/O pool. A conventional thin cylindrical shell design appears to be limited to flux levels somewhat below goal values by plate temperatures and hydraulic conditions in the long, thin coolant channels, and to be life limited by cladding oxidation. A split or ''double donut'' core configuration with much shorter heated flow paths and a coolant mixing plenum between the upper and lower core sections, appears to be capable of enhanced thermal hydraulic performance, as well as improved neutronic access. It appears that evolutionary, rather than revolutionary, advances in plate-fuel technologies will indeed be capable of producing 10/sup 16/ n/cm/sup 2/.s goal flux levels in a user friendly environment.

  6. 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.

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

    SciTech Connect

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

    1985-01-01

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

  8. 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).

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

    NASA Astrophysics Data System (ADS)

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

    2007-10-01

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

  10. 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.

  11. Evaluation of high plutonia (44% PuO 2) MOX as a fuel for fast breeder test reactor

    NASA Astrophysics Data System (ADS)

    Sengupta, A. K.; Khan, K. B.; Panakkal, Jose; Kamath, H. S.; Banerjee, S.

    2009-03-01

    Uranium plutonium mixed oxide (MOX) containing up to 30% plutonia is the conventional fuel for liquid metal cooled fast breeder reactor (LMFBR). Use of high plutonia (>30%) MOX fuel in LMFBR had been of interest but not pursued. Of late, it has regained importance for faster disposition of plutonium and also for making compact fast reactors. Some of the issues of high plutonia MOX fuels which are of concern are its chemical compatibility with liquid sodium coolant, dimensional stability and low thermal conductivity. Available literature information for MOX fuel is limited to a plutonium content of 30%. Thermodynamic assessment of mixed oxide fuels indicate that with increasing plutonia oxygen potential of the fuel increases and the fuel become more prone to chemical attack by liquid sodium coolant in case of a clad breach. In the present investigation, some of these issues of MOX fuel have been studied to evaluate this fuel for its use in fast reactor. Extensive work on the out-of-pile thermo-physical properties and fuel-coolant chemical compatibility under different simulated reactor conditions has been carried out. Results of these studies were compared with the available literature information on low plutonia MOX fuel and critically analyzed to predict in reactor behaviour of this fuel containing 44% PuO 2. The results of these out-of-pile studies have been very encouraging and helped in arriving at a suitable and achievable fuel specification for utilization of this fuel in fast breeder test reactor (FBTR). As a first step of test pin irradiation programme in FBTR, eight subassemblies of the MOX fuel are undergoing irradiation in FBTR.

  12. 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.

  13. 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

  14. 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

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

    SciTech Connect

    C. A. Wemple

    2005-09-01

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

  16. 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%.

  17. High-flux years

    SciTech Connect

    Not Available

    1992-01-01

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

  18. 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.

  19. A review of experiments and results from the transient reactor test (TREAT) facility.

    SciTech Connect

    Deitrich, L. W.

    1998-07-28

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop.

  20. 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.

  1. 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.

  2. 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.

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

    SciTech Connect

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

    1986-10-01

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

  5. 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.

  6. 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.

  7. 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

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

    NASA Technical Reports Server (NTRS)

    Castner, Raymond; Wyzykowski, John; Chiapetta, Santo

    2001-01-01

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

  9. 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.

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

    SciTech Connect

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

    1986-03-01

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

  11. 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.

  12. 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…

  13. 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.

  14. 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.

  15. 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.

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

    SciTech Connect

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

    2012-08-01

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

  17. THERMAL PERFORMANCE OF A FAST NEUTRON TEST CONCEPT FOR THE ADVANCED TEST REACTOR

    SciTech Connect

    Donna Post Guillen

    2008-06-01

    Since 1967, the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL) has provided state-of-the-art experimental irradiation testing capability. A unique design is investigated herein for the purpose of providing a fast neutron flux test capability in the ATR. This new test capability could be brought on line in approximately 5 or 6 years, much sooner than a new test reactor could be built, to provide an interim fast-flux test capability in the timeframe before a fast-flux research reactor could be built. The proposed cost for this system is approximately $63M, much less than the cost of a new fast-flux test reactor. A concept has been developed to filter out a large portion of the thermal flux component by using a thermally conductive neutron absorber block. The objective of this study is to determine the feasibility of this experiment cooling concept.

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

    NASA Astrophysics Data System (ADS)

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

    2010-05-01

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

  19. 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.

  20. 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…

  1. 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.

  2. 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.

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

    DOE PAGES

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

    2014-11-01

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

  4. 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

  5. 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.

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

    PubMed

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

    2015-10-09

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

  7. Engine Tests Using High-Sulfur Diesel Fuel

    DTIC Science & Technology

    1980-09-01

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

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

    SciTech Connect

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

    2016-05-18

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and 85Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 °C in one compact. As a result, post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating

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

    SciTech Connect

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

    2016-05-18

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and 85Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 °C in one compact. As a result, post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating

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

    DOE PAGES

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

    2016-05-18

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium,more » and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and 85Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 °C in one compact. As a result, post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers

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

    DTIC Science & Technology

    1974-08-01

    pRESSURE TRANSDUCER TEST DATA 109 4. TT2 TDMPZRATRE SENSOR For the range of inlet temperature variation of interest (-650 F to +135 0 F) the thermistor ...2. Radiation Pyrometer 79 3. Compressor Discharge Pressure Sensors 102 4. Compressor Inlet Temperature Sensor 110 5. NI, N2 and Pump Speed Sensors ...closed loop acceleration control (developed). 2. A radiation pyrometer-type turbine blade temperature sensor for use in turbine blade temperature

  12. 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.

  13. 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.

  14. 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.

  15. 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.

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

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

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

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

    DOEpatents

    Kopp, Manfred K.; Valentine, Kenneth H.

    1983-01-01

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

  18. 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.

  19. 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.

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

    NASA Astrophysics Data System (ADS)

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

    1999-03-01

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

  1. 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.

  2. 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.

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

    DOE PAGES

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

    2015-07-31

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

  4. 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.

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

    SciTech Connect

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

    2014-01-01

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

  6. 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.

  7. 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.

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

    EPA Science Inventory

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

  9. 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...

  10. Developmental design, fabrication, and test of acoustic suppressors for fans of high bypass turbofan engines

    NASA Technical Reports Server (NTRS)

    Tucker, R. H.; Nelsen, M. D.; Gregg, G. E.; Palmer, F. I.

    1974-01-01

    An analysis procedure was developed for design of acoustically treated nacelles for high bypass turbofan engines. The plan was applied to the conceptual design of a nacelle for the quiet engine typical of a 707/DC-8 airplane installation. The resultant design was modified to a test nacelle design for the NASA Lewis quiet fan. The acoustic design goal was a 10 db reduction in effective perceived fan noise levels during takoff and approach. Detailed nacelle designs were subsequently developed for both the quiet engine and the quiet fan. The acoustic design goal for each nacelle was 15 db reductions in perceived fan noise levels from the inlet and fan duct. Acoustically treated nacelles were fabricated for the quiet engine and quiet fan for testing. Performance of selected inlet and fan duct lining configurations was experimentally evaluated in a flow duct. Results of the tests show that the linings perform as designed.

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

    SciTech Connect

    Pawel, R.E.

    1988-01-01

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

  12. Neutron flux reduction programs for reactor pressure vessel

    SciTech Connect

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

    2011-07-01

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

  13. 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.

  14. Reactor Antineutrino Flux and Spectrum Shape from Daya Bay

    NASA Astrophysics Data System (ADS)

    Napolitano, Jim; Daya Bay Collaboration

    2017-01-01

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

  15. 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.

  16. 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.

  17. Neutron flux parameters at irradiation positions in the new research reactor FRM-II

    NASA Astrophysics Data System (ADS)

    Lin, Xilei; Henkelmann, Richard; Türler, Andreas; Gerstenberg, Heiko; De Corte, Frans

    2006-08-01

    The new research reactor FRM-II in Garching, Germany, was at full power 20 MW for the first time on 24th August, 2004. Since then, highly thermalized neutrons are available also for neutron activation analysis (NAA). In this report, all essential neutron flux parameters needed to calculate neutron induced reaction rates based on the Høgdahl or Westcott convention are presented for all irradiation positions in this reactor.

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

    NASA Technical Reports Server (NTRS)

    Gradl, Paul R.; Schmidt, Tim

    2016-01-01

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

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

    NASA Astrophysics Data System (ADS)

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

    1992-10-01

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

  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. Jet engine testing apparatus

    SciTech Connect

    Zweifel, T.L.

    1987-03-24

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

  2. 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)

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

    NASA Technical Reports Server (NTRS)

    Colantonio, Renato O.

    1990-01-01

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

  4. 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.

  5. 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)

  6. Improved Measurement of Reactor Flux and Spectrum at Daya Bay

    NASA Astrophysics Data System (ADS)

    Zhan, Liang; Daya Bay collaboration

    2017-09-01

    A new measurement of the reactor antineutrino flux and energy spectrum by the Daya Bay experiment is reported. With a live time of 621 days, more than 1.2 million inverse beta decay (IBD) candidates were collected by eight antineutrino detectors (ADs) deployed in two near (560 m and 600 m flux-weighted baselines) and one far (16400 m flux-weighted baseline) underground experimental halls. The IBD yield was measured and the ratio to the predicted flux using the Huber+Mueller (ILL+Vogel) model was determined to be 0.946 ± 0.020 (0.992 ± 0.021). A 2.9 σ deviation was found in the measured IBD positron energy spectrum compared to the predictions. In particular, an excess of events in the region of 4-6 MeV was found with a local significance of 4.4 σ.

  7. Preliminary Options Assessment of Versatile Irradiation Test Reactor

    SciTech Connect

    Sen, Ramazan Sonat

    2017-01-01

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

  8. 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…

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

    ERIC Educational Resources Information Center

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

    2014-01-01

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

  10. 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.

  11. 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

  12. 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

  13. Space reactor fuel element testing in upgraded TREAT

    NASA Astrophysics Data System (ADS)

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

    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 are discussed. 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.

  14. Space reactor fuel element testing in upgraded TREAT

    NASA Astrophysics Data System (ADS)

    Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

    1993-01-01

    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 ˜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 ˜100 MW/L may be achievable.

  15. 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.

  16. 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.

  17. 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

  18. 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

  19. 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)

  20. 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

  1. AJ26 engine test

    NASA Image and Video Library

    2012-06-25

    NASA engineers tested an Aerojet AJ26 rocket engine on the E-1 Test Stand at Stennis Space Center on June 25, 2012, against the backdrop of the B-1/B-2 Test Stand. The engine will be used by Orbital Sciences Corporation to power commercial cargo flights to the International Space Station.

  2. 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.

  3. 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.

  4. Development of a reactor engineering workstation at Seabrook station

    SciTech Connect

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

    1992-01-01

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

  5. 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).

  6. 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

  7. 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.

  8. 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.

  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. 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.

  11. 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)

  12. Flight Test Engineering

    NASA Technical Reports Server (NTRS)

    Pavlock, Kate Maureen

    2013-01-01

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

  13. 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.

  14. 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.

  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. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Prototype oxide breeder tests in the Fast Flux Test Facility

    SciTech Connect

    Pitner, A.L.; Gneiting, B.C.; Baker, R.B.; Hecht, S.L. )

    1994-03-01

    Four prototype irradiation tests were conducted in the Fast Flux Test Facility to investigate the performance of a 2-yr mixed-oxide fuel system using titanium-stabilized stainless steel cladding and duct material for application in a commercial-scale liquid-metal reactor plant. Three of the tests were irradiated to the point of cladding breach to establish the lifetime capability of this fuel design. Details of the fuel element design, irradiation condition and exposure, and postirradiation measurement are presented. Comparisons between measured and calculated behavior showed basically good agreement. A conservative failure analysis of the 676-fuel-pin data set from the four test assemblies indicated a 99.9% reliability for a peak burnup capability of 90 MW[center dot]/kg metal.

  1. 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.

  2. Thermal-hydraulics modeling for prototype testing of the W7-X high heat flux scraper element

    DOE PAGES

    Clark, Emily; Lumsdaine, Arnold; Boscary, Jean; ...

    2017-07-28

    The long-pulse operation of the Wendelstein 7-X (W7-X) stellarator experiment is scheduled to begin in 2020. This operational phase will be equipped with water-cooled plasma facing components to allow for longer pulse durations. Certain simulated plasma scenarios have been shown to produce heat fluxes that surpass the technological limits on the edges of the divertor target elements during steady-state operation. In order to reduce the heat load on the target elements, the addition of a “scraper element” (SE) is under investigation. The SE is composed of 24 water-cooled carbon fiber reinforced carbon composite monoblock units. Multiple full-scale prototypes have beenmore » tested in the GLADIS high heat flux test facility. Previous computational studies revealed discrepancies between the simulations and experimental measurements. In this work, single-phase thermal-hydraulics modeling was performed in ANSYS CFX to identify potential causes for such discrepancies. Possible explanations investigated were the effects of a non-uniform thermal contact resistance and a potential misalignment of the monoblock fibers. And while the difference between the experimental and computational results was not resolved by a non-uniform thermal contact resistance, the computational results provided insight into the potential performance of a W7-X monoblock unit. Circumferential temperature distributions highlighted the expected boiling regions of such a unit. Finally, simulations revealed that modest angles of fiber misalignment in the monoblocks result in asymmetries at the unit edges and provide temperature differences similar to the experimental results.« less

  3. 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.

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

    NASA Technical Reports Server (NTRS)

    Thieme, L. G.

    1981-01-01

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

  5. 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.

  6. 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.

  7. 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)

  8. High-flux isobutanol production using engineered Escherichia coli: a bioreactor study with in situ product removal.

    PubMed

    Baez, Antonino; Cho, Kwang-Myung; Liao, James C

    2011-06-01

    Promising approaches to produce higher alcohols, e.g., isobutanol, using Escherichia coli have been developed with successful results. Here, we translated the isobutanol process from shake flasks to a 1-L bioreactor in order to characterize three E. coli strains. With in situ isobutanol removal from the bioreactor using gas stripping, the engineered E. coli strain (JCL260) produced more than 50 g/L in 72 h. In addition, the isobutanol production by the parental strain (JCL16) and the high isobutanol-tolerant mutant (SA481) were compared with JCL260. Interestingly, we found that the isobutanol-tolerant strain in fact produced worse than either JCL16 or JCL260. This result suggests that in situ product removal can properly overcome isobutanol toxicity in E. coli cultures. The isobutanol productivity was approximately twofold and the titer was 9% higher than n-butanol produced by Clostridium in a similar integrated system.

  9. Preliminary Advanced Test Reactor LEU Fuel Conversion Feasibility Study

    SciTech Connect

    G. S. Chang; R. G. Ambrosek

    2005-11-01

    The Advanced Test Reactor (ATR) is a high power density, high neutron flux research reactor operating in the United States. The ATR has large irradiation test volumes located in high flux areas. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. As a result, the ATR is a representative candidate for assessing the necessary modifications and evaluating the subsequent operating effects associated with low-enriched uranium (LEU) fuel conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed for the fuel cycle burnup comparison analysis. Using the current HEU 235U enrichment of 93.0 % as a baseline, an analysis can be performed to determine the LEU uranium density and 235U enrichment required in the fuel meat to yield an equivalent Keff between the HEU core and a LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the 235U loading in the LEU core, such that the differences in Keff between the HEU and LEU core can be minimized for operation at 150 EFPD with a total core power of 115 MW. The Monte-Carlo with ORIGEN-2 (MCWO) method was used to calculate Keff versus EFPDs. The MCWO-calculated results for the LEU case demonstrated adequate excess reactivity such that the LEU core conversion designer should be able to optimize the 235U content of each fuel plate, so that the Keff and relative radial fission heat flux profile are similar to the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Upgraded Final Safety Analysis Report (UFSAR) safety requirements, a further study will be required in order to investigate the detailed radial, axial, and azimuthal heat flux profile variations versus EFPDs.

  10. 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

  11. AJ26 engine test

    NASA Image and Video Library

    2012-05-03

    NASA Deputy Administrator Lori Garver and U.S. Rep. Steven Palazzo, R-Miss., view a May 3, 2012, test of the Aerojet AJ26 rocket engine on the E-1 Test Stand at Stennis Space Center. The AJ26 engine is being tested for Orbital Sciences Corporation to power commercial cargo flights to the International Space Station.

  12. Beam choppers for neutron reflectometers at steady flux reactors

    NASA Astrophysics Data System (ADS)

    Pleshanov, N. K.

    2017-09-01

    Realizations of the TOF technique for neutron reflectometers at steady flux reactors are compared. Beam choppers for neutron reflectometers divide into choppers of type 1 (Δλ = const) and 2 (Δλ / λ = const) . It follows from Monte-Carlo simulations that choppers of type 1 do not yield to more intricate choppers of type 2, widely used at neutron reflectometers. Because of a very fast drop of neutron reflectivities with the momentum transfer q, non-optimality of measurements with a chopper of type 1 is fully compensated by better statistics at large q, and is not so much essential at small q. To vary the TOF resolution with choppers of type 1, a phasing of two discs and a turning of the system of two discs are suggested. The fluxes of neutrons with wavelengths beyond the working range and the efficiencies of their elimination by means of a bandwidth limiting prechopper are evaluated.

  13. Ultra high temperature particle bed reactor design

    NASA Astrophysics Data System (ADS)

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

    1990-04-01

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

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

    SciTech Connect

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

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

  15. AJ26 engine test

    NASA Image and Video Library

    2011-11-17

    A team of engineers at Stennis Space Center conducted a test firing of an Aerojet AJ26 flight engine Nov. 17, providing continued support to Orbital Sciences Corporation as it prepares to launch commercial cargo missions to the International Space Station. AJ26 engines will power Orbital's Taurus II rocket on the missions.

  16. 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.

  17. SP-100 reactor disassembly remote handling test program

    NASA Astrophysics Data System (ADS)

    Wilson, Charles E.; Potter, Jerry D.; Maiden, Glenn E.; Vader, David P.

    1991-01-01

    This paper is presented as an overview of the remote handling equipment validation testing, which will be conducted before installation and use in the ground engineering test facility. This equipment will be used to defuel the SP-100 reactor core after removing it from the Test Assembly following nuclear testing. A series of full scale mock-up operational tests will be conducted at a Hanford Site facility to verify equipment design, operation, and capabilities.

  18. Lessons Learned about Liquid Metal Reactors from FFTF Experience

    SciTech Connect

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

    2016-09-20

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

  19. 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.

  20. 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.

  1. The Very High Temperature Reactor

    SciTech Connect

    Hans D. Gougar; David A. Petti

    2011-06-01

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

  2. AJ26 engine test

    NASA Image and Video Library

    2011-03-19

    A team of engineers from NASA's John C. Stennis Space Center, Orbital Sciences Corporation and Aerojet conduct a successful test of an Aerojet AJ26 rocket engine on March 19. Stennis is testing AJ26 engines for Orbital Sciences to power commercial cargo missions to the International Space Station. Orbital has partnered with NASA through the Commercial Orbital Transportation Services initiative to carry out eight cargo missions to the space station by 2015, using Taurus II rockets.

  3. AJ26 engine test

    NASA Image and Video Library

    2011-12-15

    Stennis Space Center test-fired Aerojet AJ26 flight engine No. 8 on Dec. 15, continuing a commercial partnership with Orbital Services Corporation. Orbital has partnered with NASA to provide commercial cargo flights to the International Space Station. The AJ26 engines tested at Stennis will power the company's Taurus II space launch vehicle on the flights.

  4. Engineering design method for cavitational reactors: I. Sonochemical reactors

    SciTech Connect

    Gogate, P.R.; Pandit, A.B.

    2000-02-01

    High pressures and temperatures generated during the cavitation process are now considered responsible for the observed physical and chemical transformations using ultrasound irradiation. Effects of various operating parameters reported here include the frequency, the intensity of ultrasound, and the initial nuclei sizes on the bubble dynamics, and hence the magnitude of pressure generated. Rigorous solutions of the Rayleigh-Plesset equation require considerable numerical skills and the results obtained depend on various assumptions. The Rayleigh-Plesset equation was solved numerically, and the results have been empirically correlated using easily measurable global parameters in a sonochemical reactor. Liquid-phase compressibility effects were also considered. These considerations resulted in a criterion for critical ultrasound intensity, which if not considered properly can lead to overdesign or underdesign. A sound heuristic correlation, developed for the prediction of the pressure pulse generated as a function of initial nuclei sizes, frequency, and intensity of ultrasound, is valid not only over the entire range of operating parameters commonly used but also in the design procedure of sonochemical reactors with great confidence.

  5. 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

  6. 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.

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

    SciTech Connect

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

    2011-12-31

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

  8. 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.

  9. 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.

  10. Preliminary neutronic analysis of a cavity test reactor

    NASA Technical Reports Server (NTRS)

    Whitmarsh, C. L., Jr.

    1973-01-01

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

  11. 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.

  12. Preliminary Heat Transfer Characteristics of RP-2 Fuel as Tested in the High Heat Flux Facility (PREPRINT)

    DTIC Science & Technology

    2005-11-21

    2005. This facility, coupled with established heated tube facilities, such as the NASA Glenn Heated Tube Facility ( HTF ) 3 and the Wright-Patterson...operation differs from the existing facilities in both construction and heating elements. Both the NASA Glenn HTF and the Wright-Patterson Phoenix Rig...ft/sec leading to heat fluxes of up to 100 BTU/in2 sec.1 In similarity to the HTF , the HHFF uses oxygen-free grade copper walled test sections

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

    SciTech Connect

    Blanchat, Thomas K.; Hanks, Charles R.

    2013-04-01

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

  14. PRA insights applicable to the design of the Broad Applications Test Reactor

    SciTech Connect

    Khericha, S.T.; Reilly, H.J.

    1993-01-01

    Design insights applicable to the design of a new Broad Applications Test Reactor (BATR), being studied at Idaho National Engineering Laboratory, are summarized. Sources of design insights include past probabilistic risk assessments and related studies for department of Energy-owned Class A reactors and for commercial reactors. The report includes a preliminary risk allocation scheme for the BATR.

  15. 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.

  16. 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.

  17. 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.

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

    NASA Technical Reports Server (NTRS)

    1963-01-01

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

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

    NASA Technical Reports Server (NTRS)

    1963-01-01

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

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

    NASA Technical Reports Server (NTRS)

    1963-01-01

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

  1. 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.

  2. High temperature catalytic membrane reactors

    SciTech Connect

    Not Available

    1990-03-01

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

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

    SciTech Connect

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

    2012-08-15

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

  4. Impact of the neutron flux on transmutation products at fusion reactor first-walls

    NASA Astrophysics Data System (ADS)

    Sanz, J.; De La Fuente, R.; Perlado, J. M.

    1988-07-01

    To develop and to assess the suitability of a material for use as the first structural wall in a fusion reactor, it is necessary to know the transmutation behaviour of the material. In the present paper we propose a transmutation calculational strategy and how this methodology is implemented in a computer code package, called CIBELES. The code system has been developed to calculate and especially to analyze the transmutations resulting from neutron irradiation. The system includes powerful computing methods for analysing the results, and uses the numerical calculation techniques of the ORIGEN code. The transmutation characteristics of two structural materials, AISI 316L austenitic steel and DIN 1.4914 martensitic steel have been evaluated for the peripheral target position in the High Flux Isotope Reactor (HFIR), and the first wall position of the Culham Conceptual Tokamak Reactor MarkIIA (CCTRII).

  5. High Flux Heat Exchanger

    DTIC Science & Technology

    1993-01-01

    called an enhancement surfkce) is bonded to the chip. In both cases, the electronics board is immersed in a dielectric coolant . Bare chip cooling...the stocking of a dielectric coolant , which add to complexity and maintenance costs. 4. High performance heat pipe heat spreaders have demonstrated

  6. 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.

  7. 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.

  8. 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.

  9. 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

  10. 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.

  11. AJ26 engine test

    NASA Image and Video Library

    2011-02-07

    NASA Administrator Charles Bolden (l) and John C. Stennis Space Center Director Patrick Scheuermann watch the successful test of the first Aerojet AJ26 flight engine Feb. 7, 2011. The test was conducted on the E-1 Test Stand at Stennis. The engine now will be sent to Wallops Flight Facility in Virginia, where it will be used to power the first stage of Orbital Sciences Corporation's Taurus II space vehicle. The Feb. 7 test supports NASA's commitment to partner with companies to provide commercial cargo flights to the International Space Station. NASA has partnered with Orbital to carry out the first of eight cargo missions to the space station in early 2012.

  12. Savannah River Site L reactor testing

    SciTech Connect

    Menna, J.D.; Whitehouse, J.C.

    1990-01-01

    Flow tests were conducted in the Savannah River Site L reactor to evaluate the performance of the primary coolant system under simulated Loss of Coolant Accident (LOCA) conditions. Results were obtained with a prototypic cold fuel charge in the core. Core flows typical of normal and shutdown operation were studied. The tests consisted of measuring hydraulic parameters while lowering tank moderator levels to allow air entrainment from the reactor tank through operating coolant pumps. Data were collected continuously as the flows changed from single-phase to a two-component mixture of water and air. Minimum tank levels equivalent to those resulting from a hypothetical double-ended guillotine break of a coolant pipe were simulated. System pressures, water levels, densities, flows, and pump parameters were measured by over 200 instruments especially designed or adapted for in-reactor use. Special in-reactor video cameras provided visual observation of flow regimes and confirmed water levels in the reactor tank, plenum, and pump suction and plenum inlet pipes. The tests provided a unique opportunity to study full-scale pump degradation and two-component flow distributions in the reactor under ambient temperature conditions. Results showed the different pump operating regimes and points of transition and some of the other key features of the reactor response system during a severe loss of coolant event. 4 refs., 24 figs.

  13. 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.

  14. Critical heat flux test apparatus

    DOEpatents

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

    1992-01-01

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

  15. 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.

  16. AJ26 engine test

    NASA Image and Video Library

    2010-12-17

    John C. Stennis Space Center engineers conduct a 55-second test fire of Aerojet's liquid-fuel AJ26 rocket engine that will power the first stage of Orbital Sciences Corporation's Taurus II space launch vehicle. The Dec. 17, 2010 test was conducted on the E-1 Test Stand at Stennis in support of NASA's Commercial Transportation Services partnerships to enable commercial cargo flights to the International Space Station. Orbital is under contract with NASA to provide eight cargo missions to the space station through 2015.

  17. 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.

  18. 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