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Sample records for brookhaven high flux beam reactor

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

    International Nuclear Information System (INIS)

    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

  2. Job and Task Analysis project at Brookhaven National Laboratory's high flux beam reactor

    International Nuclear Information System (INIS)

    The presenter discussed the Job and Task Analysis (JTA) project conducted at Brookhaven National Laboratory's High Flux Beam Reactor (HFBR). The project's goal was to provide JTA guidelines for use by DOE contractors, then, using the guidelines conduct a JTA for the reactor operator and supervisor positions at the HFBR. Details of the job analysis and job description preparation as well as details of the task selection and task analysis were given. Post JTA improvements to the HFBR training programs were covered. The presentation concluded with a listing of the costs and impacts of the project

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

    Energy Technology Data Exchange (ETDEWEB)

    Korszun, Z.R.; Saxena, A.M.; Schneider, D.K. [Brookhaven National Laboratory, Upton, NY (United States)

    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. Technical Safety Requirement Violation at the High Flux Beam Reactor Decommissioning Project, Brookhaven, United States of America

    International Nuclear Information System (INIS)

    At Brookhaven National Laboratory (BNL) on 6 July 2009, a technical safety requirement (TSR) violation was declared at the high flux beam reactor (HFBR) project, which was a limited scope decontamination and decommissioning project associated with the permanently shutdown reactor. The violation extended from performing decommissioning activities within the facility under the incorrect mode. The draining of the spent fuel pool was performed in the warm standby mode when it should have been in the operation mode. The TSR was developed contrary to the United States Department of Energy (DOE) TSR guidance, which recommends that facility operations should only be carried out in the operation mode. The facility TSR allowed operations to be carried out in both modes. The HFBR operation mode focused on the removal of a small number of highly irradiated components with associated limited conditions of operation (LCO), while the warm standby mode focused on all other tasks in the facility and did not require entry into the LCO

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

    Science.gov (United States)

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

    1994-10-01

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

  6. Type A Verification Report For The High Flux Beam Reactor Stack And Grounds, Brookhaven National Laboratory, Upton, New York DCN: 5098-SR-08-0

    International Nuclear Information System (INIS)

    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).

  7. HFBR handbook, 1992: High flux beam reactor

    International Nuclear Information System (INIS)

    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

  8. INDEPENDENT VERIFICATION SURVEY OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT OUTSIDE AREAS BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    International Nuclear Information System (INIS)

    The objective of the verification survey was to obtain evidence by means of measurements and sampling to confirm that the final radiological conditions meet the established cleanup goals. This objective was achieved via multiple verification components including document reviews, instrument scans, and sample analysis to determine the accuracy and adequacy of FSS documentation. During the period between August 18 to 25 and September 24 to 29, 2010, ORISE conducted measurements and sampling of the HFBR 'Outside Areas' at the BNL site. ORISE performed gamma walkover scans in all eight SUs with SUs 2, 4, 6, 7, and 8 receiving high density scans of accessible areas. The remainder of SUs received low density scans. While scanning, ORISE team members observed a significant spike in count rate activity in SU 8. Just as quickly as the count rate increased the count rate decreased. A previous pass in the area did not identify any activity associated with soil contamination. The team determined that both detector instrument electronics functioned normally, and that the increased activity was due to a site activity. All individual sample concentrations and corresponding mean concentrations evaluated were determined to be below the established cleanup goal. A review of the data collected by ORISE has not identified any areas of contamination exceeding cleanup goals.

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

    Energy Technology Data Exchange (ETDEWEB)

    Harpeneau, Evan M. [Oak Ridge Institute for Science and Education, Oak Ridge, TN (United States). Independent Environmental Assessment and Verification Program

    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.

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

    International Nuclear Information System (INIS)

    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. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 5, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    International Nuclear Information System (INIS)

    The Oak Ridge Institute for Science and Education (ORISE) has reviewed the project documentation and data for the High Flux Beam Reactor (HFBR) Underground Utilities removal Phase 3; Trench 5 at Brookhaven National Laboratory (BNL) in Upton, New York. The Brookhaven Survey Group (BSG) has completed removal and performed Final Status Survey (FSS) of the concrete duct from Trench 5 from Building 801 to the Stack. Sample results have been submitted as required to demonstrate that the cleanup goal of (le)15 mrem/yr above background to a resident in 50 years has been met. Four rounds of sampling, from pre-excavation to FSS, were performed as specified in the Field Sampling Plan (FSP) (BNL 2010a). It is the policy of the U.S. Department of Energy (DOE) to perform independent verifications of decontamination and decommissioning activities conducted at DOE facilities. ORISE has been designated as the organization responsible for this task for the HFBR Underground Utilities. ORISE, together with DOE, determined that a Type A verification of Trench 5 was appropriate based on recent verification results from Trenches 2, 3, and 4, and the minimal potential for residual radioactivity in the area. The removal of underground utilities is being performed in three stages to decommission the HFBR facility and support structures. Phase 3 of this project included the removal of at least 200 feet of 36-inch to 42-inch pipe from the west side to the south side of Building 801, and the 14-inch diameter Acid Waste Line that spanned from 801 to the Stack within Trench 5. Based on the pre-excavation sample results of the soil overburden the potential for contamination of the soil surrounding the pipe is minimal (BNL 2010a). ORISE reviewed the BNL FSP and identified comments for consideration (ORISE 2010). BNL prepared a revised FSP that resolved each ORISE comment adequately (BNL 2010a). ORISE referred to the revised HFBR Underground Utilities FSP FSS data to conduct the Type A verification

  12. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 1, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    International Nuclear Information System (INIS)

    The Oak Ridge Institute for Science and Education (ORISE) has reviewed the project documentation and data for the High Flux Beam Reactor (HFBR) Underground Utilities removal Phase 3; Trench 1 at Brookhaven National Laboratory (BNL) in Upton, New York. The Brookhaven Survey Group (BSG) has completed removal and performed Final Status Survey (FSS) of the 42-inch duct and 14-inch line in Trench 1 from Building 801 to the Stack. Sample results have been submitted as required to demonstrate that the cleanup goal of (le)15 mrem/yr above background to a resident in 50 years has been met. Four rounds of sampling, from pre-excavation to FSS, were performed as specified in the Field Sampling Plan (FSP) (BNL 2010a). It is the policy of the U.S. Department of Energy (DOE) to perform independent verifications of decontamination and decommissioning activities conducted at DOE facilities. ORISE has been designated as the organization responsible for this task for the HFBR Underground Utilities. ORISE, together with DOE, determined that a Type A verification of Trench 1 was appropriate based on recent verification results from Trenches 2, 3, 4, and 5, and the minimal potential for residual radioactivity in the area. The removal of underground utilities has been performed in three stages to decommission the HFBR facility and support structures. Phase 3 of this project included the removal of at least 200 feet of 36-inch to 42-inch duct from the west side to the south side of Building 801, and the 14-inch diameter Acid Waste Line that spanned from 801 to the Stack within Trench 1. Based on the pre-excavation sample results of the soil overburden, the potential for contamination of the soil surrounding the pipe is minimal (BNL 2010a). ORISE reviewed the gamma spectroscopy results for 14 FSS soil samples, four core samples, and one duplicate sample collected from Trench 1. Sample results for the radionuclides of concern were below the established cleanup goals. However, in sample PH-3

  13. LETTER REPORT - INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT FAN HOUSE, BUILDING 704 BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    International Nuclear Information System (INIS)

    Oak Ridge Institute for Science and Education (ORISE) personnel visited the Brookhaven National Laboratory (BNL) on August 17 through August 23, 2010 to perform visual inspections and conduct independent measurement and sampling of the 'Outside Areas' at the High Flux Beam Reactor (HFBR) decommissioning project. During this visit, ORISE was also able to evaluate Fan House, Building 704 survey units (SUs) 4 and 5, which are part of the Underground Utilities portion of the HFBR decommissioning project. ORISE performed limited alpha plus beta scans of the remaining Fan House foundation lower walls and remaining pedestals while collecting static measurements. Scans were performed using gas proportional detectors coupled to ratemeter-scalers with audible output and encompassed an area of approximately 1 square meter around the static measurement location. Alpha plus beta scans ranged from 120 to 460 cpm. Twenty smears for gross alpha and beta activity and tritium were collected at judgmentally selected locations on the walls and pedestals of the Fan House foundation. Attention was given to joints, cracks, and penetrations when determining each sample location. Removable concentrations ranged from -0.43 to 1.73 dpm/100 cm2 for alpha and -3.64 to 7.80 dpm/100 cm2 for beta. Tritium results for smears ranged from -1.9 to 9.0 pCi/g. On the concrete pad, 100% of accessible area was scanned using a large area alpha plus beta gas proportional detector coupled to a ratemeter-scaler. Gross scan count rates ranged from 800 to 1500 cpm using the large area detector. Three concrete samples were collected from the pad primarily for tritium analysis. Tritium concentrations in concrete samples ranged from 53.3 to 127.5 pCi/g. Gamma spectroscopy results of radionuclide concentrations in concrete samples ranged from 0.02 to 0.11 pCi/g for Cs-137 and 0.19 to 0.22 pCi/g for Ra-226. High density scans for gamma radiation levels were performed in accessible areas in each SU, Fan House

  14. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 2 D/F WASTE LINE REMOVAL, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    International Nuclear Information System (INIS)

    Oak Ridge Institute for Science and Education (ORISE) has reviewed the project documentation and data for the High Flux Beam Reactor (HFBR) Underground Utilities removal Phase 2; the D/F Waste Line removal at Brookhaven National Laboratory (BNL) in Upton, New York. The Brookhaven Survey Group (BSG) has completed removal and performed the final status survey (FSS) of the D/F Waste Line that provided the conduit for pumping waste from Building 750 to Building 801. Sample results have been submitted as required to demonstrate that the cleanup goals of 15 mrem/yr above background to a resident in 50 years have been met. Four rounds of sampling, from pre-excavation to final status survey (FSS), were performed as specified in the Field Sampling Plan (FSP) (BNL 2010a). It is the policy of the US Departmental of Energy (DOE) to perform independent verifications of decontamination and decomissioning activities conducted at DOE facilities. ORISE has been designated as the organization responsible for this task at the HFBR. ORISE together with DOE determined that a Type A verification of the D/F Waste Line was appropriate based on its method of construction and upon the minimal potential for residual radioactivity in the area. The removal of underground utilities is being performed in three stages in the process to decommission the HFBR facility and support structures. Phase 2 of this project included the grouting and removal of 1100 feet of 2-inch pipe and 640 feet of 4-inch pipe that served as the D/F Waste Line. Based on the pre-excavation sample results of the soil overburden, the potential for contamination of the soil surrounding the pipe is minimal (BNL 2010a). ORISE reviewed the BNL FSP and identified comments for consideration (ORISE 2010). BNL prepared a revised FSP that addressed each ORISE comment adequately (BNL 2010a). ORISE referred to the revised Phase 2 D/F Waste Line removal FSP FSS data to conduct the Type A verification and determine whether the intent odf

  15. 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

    International Nuclear Information System (INIS)

    Oak Ridge Institute for Science and Education (ORISE) personnel visited the Brookhaven National Laboratory (BNL) on September 7 through September 10, 2010, and September 20 through Seeptember 24, 2010. ORISE performed visual inspections, conducted independent measurement, and sampling of Trenches 2, 3, and 4, which are part of Phase 3 for the High Flux Beam Reactor (HFBR) Underground Utilities Removal Project. Trenches 2 and 3 were addressed during the first visit and Trench 4 during the second visit to BNL. Spatial orientation to Building 801 and minimal survey area inside Trenches 2 and 3 limited satellite reception and the ability to utilize a global positioning system (GPS) as real-time data capture for the gamma scan surveys in these trenches. However, Trench 4 provided suitable conditions in which gamma scan data could be collected using the GPS. ORISE performed high-density gamma scans of accessible surface areas using shielded sodium iodide detectors coupled to ratemeter-scalers with audible output. Scans for Trench 2 ranged from 4,000 to 22,000 gross counts per minute (cpm); Trench 3 from 3,000 to 5,000 gross cpm and Trench 4 from 2,600 to 9,500 gross cpm. ORISE personnel flagged the area where the elevated counts were observed in Trench 2 for further investigation. Additional scane valuations were performed on remaining pipes and associated end-caps in the trenches with no elevated activity detected. Eleven judgemental soil samples (5098M0041 through 5098M0051) were obtained throughout Trenches 2, 3, and 4. The sample locations were selected based on count rates observed during the scan survey or because of contamination potential from pipeline removal activities. ORISE personnel judgmentally selected the location for sample M0043 in response to the 22,000 cpm observed during the scan survey, and to ascertain whether the elevataed counts were a result of soil contamination or radioactive shine from the trench's spatial orientation to the Target Room in

  16. Brookhaven leak reactor to close

    CERN Multimedia

    MacIlwain, C

    1999-01-01

    The DOE has announced that the High Flux Beam Reactor at Brookhaven is to close for good. Though the news was not unexpected researchers were angry the decision had been taken before the review to assess the impact of reopening the reactor had been concluded (1 page).

  17. Dhruva reactor -- a high flux facility for neutron beam research

    International Nuclear Information System (INIS)

    Dhruva reactor, the highest flux thermal neutron source in India has been operating at full power of 100 MW over the past two years. Several advanced facilities like the cold source, guides, etc. are being installed for neutron beam research in condensed matter. A large number and variety of neutron spectrometers are operational. This paper deals with the basic advantages that one can derive from neutron scattering investigations and gives a brief description of the instruments that are developed and commissioned at Dhruva for neutron beam research. (author). 3 figs

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

    Energy Technology Data Exchange (ETDEWEB)

    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}.

  19. Tensile and impact testing of an HFBR [High Flux Beam Reactor] control rod follower

    International Nuclear Information System (INIS)

    The Materials Technology Group of the Department of Nuclear Energy (DNE) at Brookhaven National Laboratory (BNL) undertook a program to machine and test specimens from a control rod follower from the High Flux Beam Reactor (HFBR). Tensile and Charpy impact specimens were machined and tested from non-irradiated aluminum alloys in addition to irradiated 6061-T6 from the HFBR. The tensile test results on irradiated material showed a two-fold increase in tensile strength to a maximum of 100.6 ksi. The impact resistance of the irradiated material showed a six-fold decrease in values (3 in-lb average) compared to similar non-irradiated material. Fracture toughness (KI) specimens were tested on an unirradiated compositionally and dimensionally similar (to HFBR follower) 6061 T-6 material with Kmax values of 24.8 ± 1.0 Ksi√in (average) being obtained. The report concludes that the specimens produced during the program yielded reproducible and believable results and that proper quality assurance was provided throughout the program. 9 figs., 6 tabs

  20. OPTIMIZATION OF THE EPITHERMAL NEUTRON BEAM FOR BORON NEUTRON CAPTURE THERAPY AT THE BROOKHAVEN MEDICAL RESEARCH REACTOR

    International Nuclear Information System (INIS)

    Clinical trials of Boron Neutron Capture Therapy for patients with malignant brain tumor had been carried out for half a decade, using an epithermal neutron beam at the Brookhaven's Medical Reactor. The decision to permanently close this reactor in 2000 cut short the efforts to implement a new conceptual design to optimize this beam in preparation for use with possible new protocols. Details of the conceptual design to produce a higher intensity, more forward-directed neutron beam with less contamination from gamma rays, fast and thermal neutrons are presented here for their potential applicability to other reactor facilities. Monte Carlo calculations were used to predict the flux and absorbed dose produced by the proposed design. The results were benchmarked by the dose rate and flux measurements taken at the facility then in use

  1. Neutron spectrum measurements in the aluminum oxide filtered beam facility at the Brookhaven Medical Research Reactor

    International Nuclear Information System (INIS)

    Neutron spectrum measurements were performed on the aluminum oxide filter installed in the Brookhaven Medical Research Reactor (BMRR). For these measurements, activation foils were irradiated at the exit port of the beam facility. A technique based on dominant resonances in selected activation reactions was used to measure the epithermal neutron spectrum. The fast and intermediate-energy ranges of the neutron spectrum were measured by threshold reactions and 10B-shielded 235U fission reactions. Neutron spectral data were derived from the activation data by two approaches: (1) a short analysis which yields neutron flux values at the energies of the dominant or primary resonances in the epithermal activation reactions and integral flux data for neutrons above corresponding threshold or pseudo-threshold energies, and (2) the longer analysis which utilized all the activation data in a full-spectrum, unfolding process using the FERRET spectrum adjustment code. This paper gives a brief description of the measurement techniques, analysis methods, and the results obtained

  2. Spectral characterization of the epithermal-neutron beam at the Brookhaven medical research reactor

    International Nuclear Information System (INIS)

    The power burst facility boron neutron capture therapy (PBF/BNCT) program schedule required the use of an epithermal-neutron beam before the PBF would be available. The beam was needed to carry out the acute, dose-tolerance study on healthy canines and the treatment protocol on spontaneous tumor canines. Calculations on available U.S. test reactors confirmed that the Brookhaven medical research reactor (BMRR) would be capable of providing an epithermal-neutron beam with sufficient intensity while limiting the fast-neutron and gamma dose contamination to acceptable levels for the canine irradiation studies. A joint Idaho National Engineering Laboratory (INEL)/Brookhaven National Laboratory (BNL) program was instituted to design, construct, install, and measure the performance of an epithermal-neutron beam filter for the BMRR. Aluminum oxide was selected as the filter material because it provided the desired neutron spectrum characteristics given the physical constraints of the available BMRR irradiation beam port. Neutron spectrum measurements of the exit beam were undertaken by INEL as a means to evaluate the performance of the new filter and the validity of neutron transport calculations. The preliminary data from activation measurements were presented at the Neutron Beam Design Workshop at Massachusetts Institute of Technology (MIT) in March 1989. The updated activation results and the proton-recoil measurements are presented in this paper and are compared with predictions derived from a two-dimensional transport calculation

  3. Proceedings of the Oak Ridge National Laboratory/Brookhaven National Laboratory workshop on neutron scattering instrumentation at high-flux reactors

    International Nuclear Information System (INIS)

    For the first three decades following World War II, the US, which pioneered the field of neutron scattering research, enjoyed uncontested leadership in the field. By the mid-1970's, other countries, most notably through the West European consortium at Institut Laue-Langevin (ILL) in Grenoble, France, had begun funding neutron scattering on a scale unmatched in this country. By the early 1980's, observers charged with defining US scientific priorities began to stress the need for upgrading and expansion of US research reactor facilities. The conceptual design of the ANS facility is now well under way, and line-item funding for more advanced design is being sought for FY 1992. This should lead to a construction request in FY 1994 and start-up in FY 1999, assuming an optimal funding profile. While it may be too early to finalize designs for instruments whose construction is nearly a decade removed, it is imperative that we begin to develop the necessary concepts to ensure state-of-the-art instrumentation for the ANS. It is in this context that this Instrumentation Workshop was planned. The workshop touched upon many ideas that must be considered for the ANS, and as anticipated, several of the discussions and findings were relevant to the planning of the HFBR Upgrade. In addition, this report recognizes numerous opportunities for further breakthroughs on neutron instrumentation in areas such as improved detection schemes (including better tailored scintillation materials and image plates, and increased speed in both detection and data handling), in-beam monitors, transmission white beam polarizers, multilayers and supermirrors, and more. Each individual report has been cataloged separately

  4. High Flux Isotope Reactor (HFIR)

    Data.gov (United States)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  5. Epithermal beam development at the BMRR [Brookhaven Medical Research Reactor]: Dosimetric evaluation

    International Nuclear Information System (INIS)

    The utilization of an epithermal neutron beam for neutron capture therapy (NCT) is desirable because of the increased tissue penetration relative to a thermal neutron beam. Over the past few years, modifications have been and continue to be made at the Brookhaven Medical Research Reactor (BMRR) by changing its filter components to produce an optimal epithermal beam. An optimal epithermal beam should contain a low fast neutron contamination and no thermal neutrons in the incident beam. Recently a new moderator for the epithermal beam has been installed at the epithermal port of the BMRR and has accomplished this task. This new moderator is a combination of alumina (Al2O3) bricks and aluminum (Al) plates. A 0.51 mm thick cadmium (Cd) sheet has reduced the thermal neutron intensity drastically. Furthermore, an 11.5 cm thick bismuth (Bi) plate installed at the port surface has reduced the gamma dose component to negligible levels. Foil activation techniques have been employed by using bare gold and cadmium-covered gold foil to determine thermal as well as epithermal neutron fluence. Fast neutron fluence has been determined by indium foil counting. Fast neutron and gamma dose in soft tissue, free in air, is being determined by the paired ionization chamber technique, using tissue equivalent (TE) and graphite chambers. Thermoluminescent dosimeters (TLD-700) have also been used to determine the gamma dose independently. This paper describes the methods involved in the measurements of the above mentioned parameters. Formulations have been developed and the various corrections involved have been detailed. 12 refs

  6. RELAP5/MOD2.5 analysis of the HFBR [High Flux Beam Reactor] for a loss of power and coolant accident

    International Nuclear Information System (INIS)

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

  7. Design, construction and installation of an epithermal neutron beam for BNCT at the High Flux Reactor Petten

    International Nuclear Information System (INIS)

    Following the formation in 1987, of both the European Collaboration group on Boron Neutron Capture Therapy (BNCT) and the Petten BNCT group, steps were taken to design and implement an epithermal neutron beam for BNCT applications at the High Flux Reactor (HFR) at Petten. The installation would serve as a European facility, while once the modality of BNCT is proven would be the pathfinder for implementation of BNCT at other European nuclear sites. Due to its favorable nuclear and geometric characteristics, the beam tube HB11 was chosen as the candidate beam tube for BNCT applications. To reconfigure the beam tube to produce the required epithermal neutrons, it was first necessary to remove the existing mirror system and then to install the appropriate filter materials. Due to the fixed operating schedule of the HFR, with only one long shut-down period per year during the summer weeks for maintenance and upgrading actions, installation of the new facility was planned for the summer stop period in 1990

  8. Assessment of similarity of HFBR [High Flux Beam Reactor] with separate effects test

    International Nuclear Information System (INIS)

    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 ft2), the heat flux limit obtained from the 1963 test

  9. Dose measurements and calculations in the epithermal neutron beam at the Brookhaven Medical Research Reactor (BMRR)

    International Nuclear Information System (INIS)

    The characteristics of the epithermal neutron beam at BMRR were measured, calculated, and reported by R.G. Fairchild. This beam has already been used for animal irradiations. The authors anticipate that it will be used for clinical trials. Thermal and epithermal neutron flux densities distributions, and dose rate distributions, as a function of depth were measured in a lucite dog-head phantom. Monte Carlo calculations were performed and compared with the measured values

  10. BROOKHAVEN: High energy gold

    International Nuclear Information System (INIS)

    On April 24, Brookhaven's Alternating Gradient Synchrotron (AGS) started to deliver gold ions at 11.4 GeV per nucleon (2,000 GeV per ion) to experimenters who were delighted not only to receive the world's highest energy gold beam but also to receive it on schedule

  11. Biological efficiency of the Brookhaven Medical Research Reactor mixed neutron beam estimated from gene mutations in Tradescantia stamen hair cells assay

    International Nuclear Information System (INIS)

    The relative biological effectiveness (RBE) of low energy neutrons for the induction of various abnormalities in Tradescantia stamen hair mutation (Trad-SH) assay was studied using two clones (T-4430 and T-02), heterozygous for flower color. Dose response relationship for gene mutations induced in somatic cells of Trad-SH were investigated after irradiation with a mixed neutron beam of the Brookhaven Medical Research Reactor (BMRR), currently used in a clinical trial of boron neutron capture therapy (BNCT) for glioblastoma. To establish the RBE of the BMRR beam in the induction of various biological end-points in Tradescantia, irradiation with various doses of γ-rays was also performed. After irradiation all plants were cultivated several days at Brookhaven National Laboratory (BNL), then transported to Poland for screening the biological end-points. Due to the post-exposure treatment, all plants showed high levels of lethal events and alteration of the cell cycle. Plants of clone 4430 were more reactive to post-treatment conditions, resulting in decreased blooming efficiency that affected the statistics. Slope coefficients estimated from the dose response curves for gene mutation frequencies allowed the evaluation of ranges for the maximal RBE values of the applied beam vs. γ rays as 6.0 and 5.4 for the cells of T-02 and T-4430, respectively. Estimated fraction of doses from neutrons and corresponding biological effects for the clones T-02 and T-4430 allowed to evaluate the RBE values for neutrons part in the beam as 32.3 and 45.4, respectively. (author)

  12. Nuclear structure research at the High Flux Beam Reactor: Progress report for the period February 1, 1988--September 30, 1988

    International Nuclear Information System (INIS)

    The Clark University research program in nuclear structure is a collaborative effort involving Clark University personnel, staff members from Brookhaven National Laboratory and from Ames Laboratory (Iowa State University), and an active participation of foreign scientists. The TRISTAN on-line isotope separator and the capture γ-ray facility at the HFBR are the experimental foci of the program which has four principal research themes, three involving nuclear structure physics and one directed towards astrophysics. These themes are: the manifestation of the proton-neutron interaction in the evolution of nuclear structure and its relation to collectivity, the appearance and the role of symmetries and supersymmetries in nuclei, the study of new regions of magic nuclei, and the characterization of nuclei important in r-process stellar nucleosynthesis. The activities involving Clark personnel during the eight month period, February 1, 1988 -- September 30, 1988, are summarized below

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

    International Nuclear Information System (INIS)

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

  14. The High Flux Reactor Petten, present status and prospects

    International Nuclear Information System (INIS)

    The High Flux Reactor (HFR) in Petten, The Netherlands, is a light water cooled and moderated multipurpose research reactor of the closed-tank in pool type. It is operated with highly enriched Uranium fuel at a power of 45 MW. The reactor is owned by the European Communities and operated under contract by the Dutch ECN. The HFR programme is funded by The Netherlands and Germany, a smaller share comes from the specific programmes of the Joint Research Centre (JRC) and from third party contract work. Since its first criticality in 1961 the reactor has been continuously upgraded by implementing developments in fuel element technology and increasing the power from 20 MW to the present 45 MV. In 1984 the reactor vessel was replaced by a new one with an improved accessibility for experiments. In the following years also other ageing equipment has been replaced (primary heat exchangers, pool heat exchanger, beryllium reflector elements, nuclear and process instrumentation, uninterruptable power supply). Control room upgrading is under preparation. A new safety analysis is near to completion and will form the basis for a renewed license. The reactor is used for nuclear energy related research (structural materials and fuel irradiations for LWR's, HTR's and FBR's, fusion materials irradiations). The beam tubes are used for nuclear physics as well as solid state and materials sciences. Radioisotope production at large scale, processing of gemstones and silicon with neutrons, neutron radiography and activation analysis are actively pursued. A clinical facility for boron neutron capture therapy is being designed at one of the large cross section beam tubes. It is foreseen to operate the reactor at least for a further decade. The exploitation pattern may undergo some changes depending on the requirements of the supporting countries and the JRC programmes. (author)

  15. Operation of the High Flux Reactor. Annual report 1985

    International Nuclear Information System (INIS)

    This year was characterized by the end of a major rebuilding of the installation during which the reactor vessel and its peripheral components were replaced by new and redesigned equipment. Both operational safety and experimental use were largely improved by the replacement. The reactor went back to routine operation on February 14, 1985, and has been operating without problem since then. All performance parameters were met. Other upgrading actions started during the year concerned new heat exchangers and improvements to the reactor building complex. The experimental load of the High Flux Reactor reached a satisfactory level with an average of 57%. New developments aimed at future safety related irradiation tests and at novel applications of neutrons from the horizontal beam tubes. A unique remote encapsulation hot cell facility became available adding new possibilities for fast breeder fuel testing and for intermediate specimen examination. The HFR Programme hosted an international meeting on development and use of reduced enrichment fuel for research reactors. All aspects of core physics, manufacture technology, and licensing of novel, proliferation-free, research reactor fuel were debated

  16. 1984 Operation of the high flux reactor

    International Nuclear Information System (INIS)

    The programme resources in 1984 were largely devoted to the replacement of the old reactor vessel and its peripheral equipment. The original vessel had been in operation for more than 20 years and doubts had arisen about the condition of the aluminium tank after so long an exposure to neutrons. The operation, which had never been attempted before on a reactor of that size and complexity was planned and prepared over a number of years to take advantage of the occasion to provide a much improved vessel, incorporating the latest design features. The plant was shut down at the end of November 1983 and the 14 months operation began with a short cooling-off period for decay of short lived radioactivity followed by removal of the old tank and its dissection into pieces convenient for consolidation and storage as radioactive waste. After decontamination of the shielding pool, the new vessel and neutron beam tubes were installed and the reactor was recommissioned. Routine 45 MW operation was resumed on 14 February 1985 and has been uneventful since then

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

    Science.gov (United States)

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

    2000-03-01

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

  18. Reactor operations informal montly report, May 1, 1995--May 31, 1995

    International Nuclear Information System (INIS)

    This document is an informal progress report for the operational performance of the Brookhaven Medical Research Reactor, and the Brookhaven High Flux Beam Reactor, for the month of May, 1995. Both machines ran well during this period, with no reportable instrumentation problems, all scheduled maintenance performed, and only one reportable occurance, involving a particle on Vest Button, Personnel Radioactive Contamination

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

    International Nuclear Information System (INIS)

    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 (LH2) 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 LH2 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  1. Control Rods in high-Flux Swimming-Pool Reactors

    International Nuclear Information System (INIS)

    Control-rod problems in open swimming-pool high-flux and high specific power research reactors are examined in the light of the calibrations and experiments made during the construction of the SILOE reactor. Control-rod operating experience for this reactor at 13 MW is also described. 2. The following are considered in turn: (a) Reactivity balances and reactivity values for the different types of rod tested (cadmium, B4C , rare earths and combinations of these different elements). (b) Flux peaks set up in the core by the presence of the control rods, their incidence on the specific power, the fast fluxes that can be obtained and means of increasing them. (c ) The technological problems involved in constructing the rods. (d) In-pile cooling, vibration, deformation and scram-time problems. 3. In conclusion, current studies on control rods in open swimming-pool reactors operating in the 10 - 30 1W range are briefly summarized. (author)

  2. Annual Report 1991. Operation of the high flux reactor

    International Nuclear Information System (INIS)

    In 1991 the operation of the High Flux Reactor was carried out as planned. The availability was more than 100% of scheduled operating time. The average utilization of the reactor was 69% of the practical limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons, for radioisotope production, and for various smaller activities. Development activities addressed upgrading of irradiation devices, neutron capture therapy, neutron radiography and neutron transmutation doping of silicon. General activities in support of running irradiation programmes progressed in the normal way

  3. Annual report 1989 operation of the high flux reactor

    International Nuclear Information System (INIS)

    In 1989 the operation of the High Flux Reactor Petten was carried out as planned. The availability was more than 100% of scheduled operating time. The average occupation of the reactor by experimental devices was 72% of the practical occupation limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons and for radioisotope production. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

  4. Annual report 1990. Operation of the high flux reactor

    International Nuclear Information System (INIS)

    In 1990 the operation of the High Flux Reactor was carried out as planned. The availability was 96% of scheduled operating time. The average utilization of the reactor was 71% of the practical limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons, for radioisotope production, and for various smaller activities. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

  5. RELAP5 model of the high flux isotope reactor with low enriched fuel thermal flux profiles

    Energy Technology Data Exchange (ETDEWEB)

    Banfield, J.; Mervin, B.; Hart, S.; Ritchie, J.; Walker, S.; Ruggles, A.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee Knoxville, Knoxville, TN 37996-2300 (United States)

    2012-07-01

    The High Flux Isotope Reactor (HFIR) currently uses highly enriched uranium (HEU) fabricated into involute-shaped fuel plates. It is desired that HFIR be able to use low enriched uranium (LEU) fuel while preserving the current performance capability for its diverse missions in material irradiation studies, isotope production, and the use of neutron beam lines for basic research. Preliminary neutronics and depletion simulations of HFIR with LEU fuel have arrived to feasible fuel loadings that maintain the neutronics performance of the reactor. This article illustrates preliminary models developed for the analysis of the thermal-hydraulic characteristics of the LEU core to ensure safe operation of the reactor. The beginning of life (BOL) LEU thermal flux profile has been modeled in RELAP5 to facilitate steady state simulation of the core cooling, and of anticipated and unanticipated transients. Steady state results are presented to validate the new thermal power profile inputs. A power ramp, slow depressurization at the outlet, and flow coast down transients are also evaluated. (authors)

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

    International Nuclear Information System (INIS)

    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 LH2 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

  7. Brookhaven highlights

    International Nuclear Information System (INIS)

    This publication provides a broad overview of the research programs and efforts being conducted, built, designed, and planned at Brookhaven National Laboratory. This work covers a broad range of scientific disciplines. Major facilities include the Alternating Gradient Synchrotron (AGS), with its newly completed booster, the National Synchrotron Light Source (NSLS), the High Flux Beam Reactor (HFBR), and the RHIC, which is under construction. Departments within the laboratory include the AGS department, accelerator development, physics, chemistry, biology, NSLS, medical, nuclear energy, and interdepartmental research efforts. Research ranges from the pure sciences, in nuclear physics and high energy physics as one example, to environmental work in applied science to study climatic effects, from efforts in biology which are a component of the human genome project to the study, production, and characterization of new materials. The paper provides an overview of the laboratory operations during 1992, including staffing, research, honors, funding, and general laboratory plans for the future

  8. Brookhaven highlights

    Energy Technology Data Exchange (ETDEWEB)

    Rowe, M.S.; Cohen, A.; Greenberg, D.; Seubert, L. (eds.)

    1992-01-01

    This publication provides a broad overview of the research programs and efforts being conducted, built, designed, and planned at Brookhaven National Laboratory. This work covers a broad range of scientific disciplines. Major facilities include the Alternating Gradient Synchrotron (AGS), with its newly completed booster, the National Synchrotron Light Source (NSLS), the High Flux Beam Reactor (HFBR), and the RHIC, which is under construction. Departments within the laboratory include the AGS department, accelerator development, physics, chemistry, biology, NSLS, medical, nuclear energy, and interdepartmental research efforts. Research ranges from the pure sciences, in nuclear physics and high energy physics as one example, to environmental work in applied science to study climatic effects, from efforts in biology which are a component of the human genome project to the study, production, and characterization of new materials. The paper provides an overview of the laboratory operations during 1992, including staffing, research, honors, funding, and general laboratory plans for the future.

  9. Transverse beam dampers for the Brookhaven AGS

    International Nuclear Information System (INIS)

    A wide band damper system has been developed for the Brookhaven Alternating Gradient Synchrotron (AGS). The system consists of two sets of PUE pickups, analog and digital processing electronics, four 500 Watt wide band power amplifiers, and two pairs of strip line deflectors. The system is currently used to damp transverse coherent instabilities and injection errors, in both planes, for protons and all species of Heavy Ions. This paper discusses the system design and operation, focusing on the engineering considerations and problems encountered in the actual implementation. Operational data from both protons and Heavy Ion beams is presented

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bryant, Rebecca [Bryant Research, LLC; Kszos, Lynn A [ORNL

    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

  12. High flux testing reactor Petten. Replacement of the reactor vessel and connected components. Overall report

    International Nuclear Information System (INIS)

    The project of replacing the HFR originated in 1974 when results of several research programmes confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report contains the detailed chronology of events concerning preparation and execution of the replacement. After a 14 months' outage the reactor resumed routine operation on 14th February, 1985. At the end of several years of planning and preparation the reconstruction proceded in the following steps: unloading of the old core, decay of short-lived radioactivity in December 1983, removal of the old tank and of its peripheral equipment in January-February 1984, segmentation and waste disposal of the removed components in March-April, decontamination of the pools, bottom penetration overhauling in May-June, installation of the new tank and other new components in July-September, testing and commissioning, including minor modifications in October-December, and, trials runs and start-up preparation in January-February 1985. The new HFR Petten features increased and improved experimental facilities. Among others the obsolete thermal columns was replaced by two high flux beam tubes. Moreover the new plant has been designed for future increases of reactor power and neutron fluxes. For the next three to four years the reactor has to cope with a large irradiation programme, claiming its capacity to nearly 100%

  13. Annual progress report 1988, operation of the high flux reactor

    International Nuclear Information System (INIS)

    In 1988 the High Flux Reactor Petten was routinely operated without any unforeseen event. The availability was 99% of scheduled operation. Utilization of the irradiation positions amounted to 80% of the practical occupation limit. The exploitation pattern comprised nuclear energy deployment, fundamental research with neutrons, and radioisotope production. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

  14. Design of small-animal thermal neutron irradiation facility at the Brookhaven Medical Research Reactor

    International Nuclear Information System (INIS)

    The broad beam facility (BBF) at the Brookhaven Medical Research Reactor (BMRR) can provide a thermal neutron beam with flux intensity and quality comparable to the beam currently used for research on neutron capture therapy using cell-culture and small-animal irradiations. Monte Carlo computations were made, first, to compare with the dosimetric measurements at the existing BBF and, second, to calculate the neutron and gamma fluxes and doses expected at the proposed BBF. Multiple cell cultures or small animals could be irradiated simultaneously at the so-modified BBF under conditions similar to or better than those individual animals irradiated at the existing thermal neutron irradiation Facility (TNIF) of the BMRR. The flux intensity of the collimated thermal neutron beam at the proposed BBF would be 1.7 x 1010 n/cm2·s at 3-MW reactor power, the same as at the TNIF. However, the proposed collimated beam would have much lower gamma (0.89 x 10-11 cGy·cm2/nth) and fast neutron (0.58 x 10-11 cGy·cm2/nth) contaminations, 64 and 19% of those at the TNIF, respectively. The feasibility of remodeling the facility is discussed

  15. Fuel management at the Petten high flux reactor

    International Nuclear Information System (INIS)

    Several years ago the shipment of spent fuel of the High Flux Reactor (HFR) at Petten has come to a standstill resulting in an ever growing stock of fuel elements that are labelled 'fully burnt up'. Examination of those elements showed that a reasonably number of them have a relatively high 235U mass left. A reactor physics analysis showed that the use of such elements in the peripheral core zone allows the loading of four instead of five fresh fuel elements in many cycle cores. For the assessment of safety and performance parameters of HFR cores a new calculational tool is being developed. It is based on AEA Technology's Reactor physics code suite Winfrith Improved Multigroup Scheme (WIMS). NRG produced pre- and post-processing facilities to feed input data into WIMS's 2D transport code CACTUS and to extract relevant parameters from the output. The processing facilities can be used for many different types of application. (author)

  16. Development of the epithermal neutron beam and its clinical application for boron neutron capture therapy at the Brookhaven medical research reactor

    International Nuclear Information System (INIS)

    The failures of the Boron Neutron Capture Therapy (BNCT) trials conducted between 1951 and 1961 were attributed to inadequate penetration of the thermal neutron beams and poor localization of boron compound in the tumour. The epithermal neutron beam at the BMRR was designed and installed to improve the penetration of the neutron beam. The use of this epithermal neutron beam for the clinical trial initiated in 1994 at Brookhaven National Laboratory (BNL) was preceded by the neutron beam optimization and characterization, the validation of the treatment planning software and the establishment of a procedure for treatment plan evaluation and dose reporting and recording. To date, a total of 54 patients have been treated. Our experience in the development of the epithermal neutron beam for clinical BNCT at the BMRR may be useful to other investigators desirous of developing similar programs for cancer therapy. (author)

  17. Surveillance programme and upgrading of the High Flux Reactor Petten

    International Nuclear Information System (INIS)

    The High Flux Reactor (HFR) at Petten (The Netherlands), a 45 MW light water cooled and moderated research reactor in operation during more than 30 years, has been kept up to date by replacing ageing components. In 1984, the HFR was shut down for replacement of the aluminium. reactor vessel which had been irradiated during more than 20 years. The demonstration that the new vessel contains no critical defect requires knowledge of the material properties of the aluminium alloy Al 5154 with and without neutron irradiation and of the likely defect presence through the periodic in-service inspections. An irradiation damage surveillance programme has been started in 1985 for the new vessel material to provide information on fracture mechanics properties. After the vessel replacement, the existing process of continuous upgrading and replacement of ageing components was accelerated. A stepwise upgrade of the control room is presently under realization. (author)

  18. Brookhaven highlights, fiscal year 1985, October 1, 1984-September 30, 1985

    Energy Technology Data Exchange (ETDEWEB)

    1985-01-01

    Activities at Brookhaven National Laboratory are briefly discussed. These include work at the National Synchrotron Light Source, the High Flux Beam Reactor, and the Alternating Gradient Synchrotron. Areas of research include heavy ion reactions, neutrino oscillations, low-level waste, nuclear data, medicine, biology, chemistry, parallel computing, optics. Also provided are general and administrative news, a financial report. (LEW)

  19. Brookhaven highlights, fiscal year 1985, October 1, 1984-September 30, 1985

    International Nuclear Information System (INIS)

    Activities at Brookhaven National Laboratory are briefly discussed. These include work at the National Synchrotron Light Source, the High Flux Beam Reactor, and the Alternating Gradient Synchrotron. Areas of research include heavy ion reactions, neutrino oscillations, low-level waste, nuclear data, medicine, biology, chemistry, parallel computing, optics. Also provided are general and administrative news, a financial report

  20. Neutronics modeling of the High Flux Isotope Reactor using COMSOL

    International Nuclear Information System (INIS)

    Highlights: → Neutron flux distributions in HFIR were calculated with SCALE v6 and COMSOL v3.5. → Diffusion theory employed in COMSOL coefficient partial differential equation mode. → Two-group 2D flux distributions compare well to benchmarked 3D stochastic models. → Adaptive mesh refinement algorithm used to refine mesh and improve accuracy. → First step in a long-term project to upgrade research reactor analytical methods. - Abstract: 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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. Physics design for the Brookhaven Medical Research Reactor epithermal neutron source

    International Nuclear Information System (INIS)

    A collaborative effort by researchers at the Idaho National Engineering Laboratory and the Brookhaven National Laboratory has resulted in the design and implementation of an epithermal-neutron source at the Brookhaven Medical Research Reactor (BMRR). Large aluminum containers, filled with aluminum oxide tiles and aluminum spacers, were tailored to pre-existing compartments on the animal side of the reactor facility. A layer of cadmium was used to minimize the thermal-neutron component. Additional bismuth was added to the pre-existing bismuth shield to minimize the gamma component of the beam. Lead was also added to reduce gamma streaming around the bismuth. The physics design methods are outlined in this paper. Information available to date shows close agreement between calculated and measured beam parameters. The neutron spectrum is predominantly in the intermediate energy range (0.5 eV - 10 keV). The peak flux intensity is 6.4E + 12 n/(m2.s.MW) at the center of the beam on the outer surface of the final gamma shield. The corresponding neutron current is 3.8E + 12 n/(m2.s.MW). Presently, the core operates at a maximum of 3 MW. The fast-neutron KERMA is 3.6E-15 cGy/(n/m2) and the gamma KERMA is 5.0E-16 cGY/(n/m2) for the unperturbed beam. The neutron intensity falls off rapidly with distance from the outer shield and the thermal flux realized in phantom or tissue is strongly dependent on the beam-delimiter and target geometry

  3. A Compact, High-Flux Cold Atom Beam Source

    Science.gov (United States)

    Kellogg, James R.; Kohel, James M.; Thompson, Robert J.; Aveline, David C.; Yu, Nan; Schlippert, Dennis

    2012-01-01

    The performance of cold atom experiments relying on three-dimensional magneto-optical trap techniques can be greatly enhanced by employing a highflux cold atom beam to obtain high atom loading rates while maintaining low background pressures in the UHV MOT (ultra-high vacuum magneto-optical trap) regions. Several techniques exist for generating slow beams of cold atoms. However, one of the technically simplest approaches is a two-dimensional (2D) MOT. Such an atom source typically employs at least two orthogonal trapping beams, plus an additional longitudinal "push" beam to yield maximum atomic flux. A 2D atom source was created with angled trapping collimators that not only traps atoms in two orthogonal directions, but also provides a longitudinal pushing component that eliminates the need for an additional push beam. This development reduces the overall package size, which in turn, makes the 2D trap simpler, and requires less total optical power. The atom source is more compact than a previously published effort, and has greater than an order of magnitude improved loading performance.

  4. A high energy neutral beam system for reactors

    International Nuclear Information System (INIS)

    High energy neutral beams provide a promising method of heating and driving current in steady-stage tokamak fusion reactors. As an example, we have made a conceptual design of a neutral beam system for current drive on the International Thermonuclear Experimental Reactor (ITER). The system, based on electrostatic acceleration of Dions, can deliver up to 100 MW of 1.6 MeV Do neutrals through three ports. Radiation protection is provided by locating sensitive beamlime components 35 to 50 m from the reactor. In an application to a 3300 MW power reactor, a system delivering 120 MW of 2-2.4 MeV deuterium beams assisted by 21 MW of lower hybrid wave power drives 25 MA provides an adequate plasma power again (Q = 24) for a commercial fusion power plant. (author). 8 refs.; 1 fig.; 2 tabs

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Evan Harpeneau

    2011-06-24

    The Separations Process Research Unit (SPRU) complex located on the Knolls Atomic Power Laboratory (KAPL) site in Niskayuna, New York, was constructed in the late 1940s to research the chemical separation of plutonium and uranium (Figure A-1). SPRU operated as a laboratory scale research facility between February 1950 and October 1953. The research activities ceased following the successful development of the reduction oxidation and plutonium/uranium extraction processes. The oxidation and extraction processes were subsequently developed for large scale use by the Hanford and Savannah River sites (aRc 2008a). Decommissioning of the SPRU facilities began in October 1953 and continued through the 1990s.

  8. Final Report Independent Verification Survey of the High Flux Beam Reactor, Building 802 Fan House Brookhaven National Laboratory Upton, New York DCN: 5098-SR-06-0

    International Nuclear Information System (INIS)

    The Separations Process Research Unit (SPRU) complex located on the Knolls Atomic Power Laboratory (KAPL) site in Niskayuna, New York, was constructed in the late 1940s to research the chemical separation of plutonium and uranium (Figure A-1). SPRU operated as a laboratory scale research facility between February 1950 and October 1953. The research activities ceased following the successful development of the reduction oxidation and plutonium/uranium extraction processes. The oxidation and extraction processes were subsequently developed for large scale use by the Hanford and Savannah River sites (aRc 2008a). Decommissioning of the SPRU facilities began in October 1953 and continued through the 1990s.

  9. On the cycle and neutron fluxes of the high flux reactor at ill Grenoble

    International Nuclear Information System (INIS)

    The fuel evolution and neutron fluxes of the high flux reactor at ILL Grenoble were investigated for the full fuel cycle. Two different code systems, namely MONTEBURNS and MCNP + CINDER'90, were employed for a realistic 3-D geometry description of the reactor core and experimental channels of interest. Both codes correctly reproduce the history of the fuel burn-up. Calculated neutron flux in the reactor core is 2.16 x 1015 n/s cm2 gym- and decreases down to 1.65 x 1015 n/s cm2 - and 6.7 x 1014 n/s cm2 in the experimental channels V4 and H9 respectively. Thermal neutron contribution is ∼85.2% in V4 and ∼98.3% in H9. We show that neutron fluxes in the experimental channels V4 and H9 are not perturbed/changed due to the burn-up of the fuel element and/or movement of the control rod. This result should simplify most of the irradiation experiments (and data analysis in particular), which employ the experimental channels as above. (authors)

  10. Testing of research reactor fuel in the high flux reactor (Petten)

    International Nuclear Information System (INIS)

    The two types of fuel most frequently used by the main research reactors are metallic: highly enriched uranium (>90%) and silicide low enriched uranium (3. However, a need exists for research on new reactor fuel. This would permit some plants to convert without losses in flux or in cycle length and would allow new reactor projects to achieve higher possibilities especially in fluxes. In these cases research is made either on silicide with higher density, or on other types of fuel (UMo, etc.). In all cases when new fuel is proposed, there is a need, for safety reasons, to test it, especially regarding the mechanical evolution due to burn-up (swelling, etc.). Initially, such tests are often made with separate plates, but lately, using entire elements. Destructive examinations are often necessary. For this type of test, the High Flux Reactor, located in Petten (The Netherlands) has many specific advantages: a large core, providing a variety of interesting positions with high fluence rate; a downward coolant flow simplifies the engineering of the device; there exists easy access with all handling possibilities to the hot-cells; the high number of operating days (>280 days/year), together with the high flux, gives a possibility to reach quickly the high burn-up needs; an experienced engineering department capable of translating specific requirements to tailor-made experimental devices; a well equipped hot-cell laboratory on site to perform all necessary measurements (swelling, γ-scanning, profilometry) and all destructive examinations. In conclusion, the HFR reactor readily permits experimental research on specific fuels used for research reactors with all the necessary facilities on the Petten site. (author)

  11. Safety and quality management at the high flux reactor Petten

    International Nuclear Information System (INIS)

    The High Flux Reactor (HFR) is one high power multi-purpose materials testing research reactor of the tank-in-pool type, cooled and moderated by light-water. It is operated at 45 MW at a prescribed schedule of 11 cycles per year, each comprising 25 operation days and three shut-down days. Since the licence for the operation of HFR was granted in 1962, a total of 14 amendments to the original licence have been made following different modifications in the installations. In the meantime, international nuclear standards were developed, especially in the framework of the NUSS programme of the IAEA, which were adopted by the Dutch Licensing Authorities. In order to implement the new standards, the situation at the HFR was comprehensively reviewed in the course of an audit performed by the Dutch Licensing Authorities in 1988. This also resulted in formulating the task of setting-up an 'HFR - Integral Quality Assurance Handbook' (HFR-IQAD) involving both organizations JRCIAM and ECN, which had the unique framework and basic guideline to assure the safe and efficient operation and exploitation of the HFR and to promote safety and quality in all aspects of HFR related activities. The assurance of safe and efficient operation and exploitation of the HFR is condensed together under the concepts of safety and quality of services and is achieved through the safety and quality management. (orig.)

  12. High Flux Isotope Reactor system RELAP5 input model

    International Nuclear Information System (INIS)

    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

  13. High Flux Isotope Reactor system RELAP5 input model

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Overview of irradiation facilities and experiments currently in the Oak Ridge High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    The Oak Ridge High Flux Isotope Reactor (HFIR) is an 85 MW research reactor with a variety of irradiation facilities. The target region has the highest continuous thermal neutron flux available in the western world and facilities in the beryllium reflector provide opportunities to irradiate experiments of various sizes in a variety of neutron spectrums. Major programs utilizing these facilities include Fusion Materials, Advanced Neutron Source (ANS), New Production Reactor, and Modular High Temperature Gas-Cooled Reactor

  15. Alize 3 - first critical experiment for the franco-german high flux reactor - calculations

    International Nuclear Information System (INIS)

    The results of experiments in the light water cooled D2O reflected critical assembly ALIZE III have been compared to calculations. A diffusion model was used with 3 fast and epithermal groups and two overlapping thermal groups, which leads to good agreement of calculated and measured power maps, even in the case of strong variations of the neutron spectrum in the core. The difference of calculated and measured keff was smaller than 0.5 per cent δk/k. Calculations of void and structure material coefficients of the reactivity of 'black' rods in the reflector, of spectrum variations (Cd-ratio, Pu-U-ratio) and to the delayed photoneutron fraction in the D2O reflector were made. Measurements of the influence of beam tubes on reactivity and flux distribution in the reflector were interpreted with regard to an optimum beam tube arrangement for the Franco- German High Flux Reactor. (author)

  16. Extraction of gadolinium from high flux isotope reactor control plates

    International Nuclear Information System (INIS)

    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 153Gd 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 153Gd, 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 153Gd. Based on the results of experimental and calculational analyses, up to 25 grams of valuable gadolinium (≥60% enriched in 152Gd) 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 153Gd for each gram of gadolinium re-irradiated, 700 one-curie sources can be produced from each control plate assayed

  17. Neutron scattering at the high-flux isotope reactor

    International Nuclear Information System (INIS)

    The title facilities offer the brightest source of neutrons in the national user program. Neutron scattering experiments probe the structure and dynamics of materials in unique and complementary ways as compared to x-ray scattering methods and provide fundamental data on materials of interest to solid state physicists, chemists, biologists, polymer scientists, colloid scientists, mineralogists, and metallurgists. Instrumentation at the High- Flux Isotope Reactor includes triple-axis spectrometers for inelastic scattering experiments, a single-crystal four diffractometer for crystal structural studies, a high-resolution powder diffractometer for nuclear and magnetic structure studies, a wide-angle diffractometer for dynamic powder studies and measurements of diffuse scattering in crystals, a small-angle neutron scattering (SANS) instrument used primarily to study structure-function relationships in polymers and biological macromolecules, a neutron reflectometer for studies of surface and thin-film structures, and residual stress instrumentation for determining macro- and micro-stresses in structural metals and ceramics. Research highlights of these areas will illustrate the current state of neutron science to study the physical properties of materials

  18. Development of a High Flux Research Reactor with Assembly- Integral Flux Traps

    International Nuclear Information System (INIS)

    In recent years, the demand for irradiation services has increased as research reactors face a number of challenges. These challenges include reduction in enrichment, the need for economical use of fuel, and the medical isotope crisis brought about by the temporary shutdowns of the NRU and HFR. These challenges have inspired the design of a revolutionary research reactor assembly concept making use of the new U-Mo monolithic fuel form. Each assembly consists of a trapezoidal plate-fueled region and integral triangular flux trap. When placed in a hexagonal core, this structure permits multiple orientations of the flux trap with respect to those surrounding it. As a result, multiple orientations, sizes and shapes of the flux trap are possible. These flux trap parameters can be adjusted to provide the spectra needed to meet ever-changing experimental needs. This assembly concept was applied to the development of a 5 MWth research reactor core, and optimized for peak thermal flux. During the design process, attention was given to making the design economical, manufacturable, and maintainable. The design was able to produce a maximum unperturbed thermal flux of 1.2 x 1014 neutrons/cm2 s, 70% higher than existing research reactors of the same power rating. Despite this optimization for thermal flux, the reactor is still capable of producing fast fluxes in excess of 1 x 1014 neutrons/cm2 s in other regions of the core. The response of the resulting core to thermal hydraulic transients was also analyzed. (author)

  19. Digital transverse beam dampers from the Brookhaven AGS

    International Nuclear Information System (INIS)

    A wide band, digital damper system has been developed and is in use at the Brookhaven Alternating Gradient Synchrotron (AGS). The system consists of vertical and horizontal capacitive pickups, analog and digital processing electronics, four 500 Watt wide band power amplifiers, and two pairs of strip line beam kickers. The system is currently used to damp transverse coherent instabilities and injection errors, in both planes, for protons and all species of heavy ions. This paper discusses the system design and operation, particularly with regard to stabilization of the high intensity proton beam. The analog and digital signal processing techniques used to achieve optimum results are discussed. Operational data showing the effect of the damping are presented

  20. Reactor vessel dismantling at the high flux materials testing reactor Petten

    International Nuclear Information System (INIS)

    The project of replacing the reactor vessel of the high flux materials testing reactor (HFR) originated in 1974 when results of several research programs confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report describes the dismantling philosophy and organisation, the design of special underwater equipment, the dismantling of the reactor vessel and thermal column, and the conditioning and shielding activities resulting in a working area for the installation of the new vessel with no access limitations due to radiation. Finally an overview of the segmentation, waste disposal and radiation exposure is given. The total dismantling, segmentation and conditioning activities resulted in a total collective radiation dose of 300 mSv. (orig.)

  1. Spectrum and density of neutron flux in the irradiation beam line no. 3 of the IBR-2 reactor

    Science.gov (United States)

    Shabalin, E. P.; Verkhoglyadov, A. E.; Bulavin, M. V.; Rogov, A. D.; Kulagin, E. N.; Kulikov, S. A.

    2015-03-01

    Methodology and results of measuring the differential density of the neutron flux in irradiation beam line no. 3 of the IBR-2 reactor using neutron activation analysis (NAA) are presented in the paper. The results are compared to the calculation performed on the basis of the 3D MCNP model. The data that are obtained are required to determine the integrated radiation dose of the studied samples at various distances from the reactor.

  2. Evaluation of HFIR [High Flux Isotope Reactor] pressure-vessel integrity considering radiation embrittlement

    International Nuclear Information System (INIS)

    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 ∼104 less), that is, a rate effect

  3. Temperature measurements during high flux ion beam irradiations.

    Science.gov (United States)

    Crespillo, M L; Graham, J T; Zhang, Y; Weber, W J

    2016-02-01

    A systematic study of the ion beam heating effect was performed in a temperature range of -170 to 900 °C using a 10 MeV Au(3+) ion beam and a Yttria stabilized Zirconia (YSZ) sample at a flux of 5.5 × 10(12) cm(-2) s(-1). Different geometric configurations of beam, sample, thermocouple positioning, and sample holder were compared to understand the heat/charge transport mechanisms responsible for the observed temperature increase. The beam heating exhibited a strong dependence on the background (initial) sample temperature with the largest temperature increases occurring at cryogenic temperatures and decreasing with increasing temperature. Comparison with numerical calculations suggests that the observed heating effect is, in reality, a predominantly electronic effect and the true temperature rise is small. A simple model was developed to explain this electronic effect in terms of an electrostatic potential that forms during ion irradiation. Such an artificial beam heating effect is potentially problematic in thermostated ion irradiation and ion beam analysis apparatus, as the operation of temperature feedback systems can be significantly distorted by this effect. PMID:26931879

  4. Temperature measurements during high flux ion beam irradiations

    Science.gov (United States)

    Crespillo, M. L.; Graham, J. T.; Zhang, Y.; Weber, W. J.

    2016-02-01

    A systematic study of the ion beam heating effect was performed in a temperature range of -170 to 900 °C using a 10 MeV Au3+ ion beam and a Yttria stabilized Zirconia (YSZ) sample at a flux of 5.5 × 1012 cm-2 s-1. Different geometric configurations of beam, sample, thermocouple positioning, and sample holder were compared to understand the heat/charge transport mechanisms responsible for the observed temperature increase. The beam heating exhibited a strong dependence on the background (initial) sample temperature with the largest temperature increases occurring at cryogenic temperatures and decreasing with increasing temperature. Comparison with numerical calculations suggests that the observed heating effect is, in reality, a predominantly electronic effect and the true temperature rise is small. A simple model was developed to explain this electronic effect in terms of an electrostatic potential that forms during ion irradiation. Such an artificial beam heating effect is potentially problematic in thermostated ion irradiation and ion beam analysis apparatus, as the operation of temperature feedback systems can be significantly distorted by this effect.

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

    International Nuclear Information System (INIS)

    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 principle' of the design has been obtained from the internal ORNL safety review committees and the U.S. 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 MCNP1 A D 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 ATHENA2 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. Our present schedule shows the assembling of the cold source loop on site 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. (author)

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

    International Nuclear Information System (INIS)

    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 (∼47%), wood (∼38%) and asbestos transite (∼14%). The remaining ∼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, except for the

  7. High resolution powder neutron diffraction at a low-flux reactor

    International Nuclear Information System (INIS)

    Complete text of publication follows. The high resolution powder neutron diffractometer, PUS, was installed at the 2 MW JEEP II reactor at Institute for Energy Technology, Kjeller in 1997. Flexibility and optimisation of the relationship between intensity (signal/noise ratio) and resolution were keywords for the design of the instrument [1]. Wavelengths in the range 0.7 - 2.6 A are available from the focussing composite Ge monochromator. Different Soller slit collimators in the incident neutron beam allow either a high resolution or a high intensity mode. The detector system consists of two units position units, each covering 20 deg in 2Θ, and a complete 120 degrees powder pattern are collected from measurements with the detector-units in three different positions. The presentation focus is on some selected examples, like crystal and magnetic structures of metal hydrides, high TC-superconductors and perovskite related oxides, to show the potential of high resolution powder neutron diffraction at a low-flux reactor, like the JEEP II reactor at Kjeller. (author) [1] B.C. Hauback, H. Fjellvag, O. Steinsvoll, K. Johansson, O.T. Buset and J. Jorgensen, to be submitted (1999)

  8. COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

    OpenAIRE

    Phani Kumar Domalapally; Slavomir Entler

    2015-01-01

    Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

  9. Status of research reactor fuel test in the High Flux Reactor (Petten)

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J.; Casalta, S. [European Commission, JRC, NL-1755 ZG Petten (Netherlands); Wijtsma, F.J.; Thijssen, P.J.M.; Hendriks, J.A.; Dassei, G. [NRG Petten (Netherlands); Vacelet, H. [Cerca Framatome, F-26104 Romans (France); Languille, A. [CEA Cadarache, F-13108 Saint Paul Lez Durance (France)

    2001-07-01

    Even if the research reactors are using very well known MTR-fuel, a need exists for research in this field mainly for the reasons of industrial qualification of fuel assemblies (built with qualified fuel), improvement or modifications on a qualified fuel ( e.g. increase of density), and qualification of a new fuels such as UMo. For these types of tests, the High Flux Reactor located in Petten (the Netherlands) has a lot of specific advantages: 1) a large core with various interesting positions ranging from high to low fluence rate; 2) a high number of operating days (>280 days/year) that gives - with the high flux available - a possibility to reach quickly high burnup; 3) a downward coolant flow that simplifies the device engineering; 4) all possibilities of non-destructive and destructive examinations in the hot-cells (visual inspection, swelling, {gamma}-scanning, macro- and light microscopy, SEM and EPMA examinations, tomography). Two types of tests can be performed at the plant: either a full-scale test or a test of plates in dedicated devices. A presentation is made of the irradiation test on four UMo plates, begun in March 2000 in the device UMUS. A status report is provided of the full-scale test to be done in the near future, especially the UMo tests to begin the next year. In conclusion it appears that the HFR, that had already given an excellent contribution to silicide fuel qualification in the 1980s, will also give a significant contribution to the current UMo qualification programs. (author)

  10. Trends and techniques in neutron beam research for medium and low flux research reactors. Report of a consultants meeting

    International Nuclear Information System (INIS)

    The IAEA is making concerted efforts to promote R and D programmes for neutron beam research to assist the developing Member States in better utilization of their research reactors. A consultants meeting was organized on 16-19 March 1996 to review the current status and deliberate on the future trends in neutron beam based research using low and medium flux research reactors with the flux range of the order of up to 1013-1014 n/cm2/s, particularly in the light of recent advances in electronics and instrumentation. The participants focused on five specific topics: triple axis spectrometry, neutron depolarization studies, capillary optics, spin-echo spectrometry and small-angle neutron spectrometry. This TECDOC details the highlights of the discussions in the meeting along with the papers presented

  11. BROOKHAVEN

    International Nuclear Information System (INIS)

    Neutrino physics has been an integral part of the Brookhaven research programme for much of the Laboratory's 46-year history. Milestones have been the determination of the helicity of neutrinos (1958), the establishment of the existence of two kinds of neutrinos (1962) for which Leon Lederman, Mel Schwartz and Jack Steinberger were awarded the 1988 Nobel Physics Prize, the discovery of charmed baryons in the 7' Bubble Chamber in 1975, and the ongoing measurements by Ray Davis and collaborators of solar neutrinos, first reported in 1968. There have also been significant contributions to the understanding of neutral currents in exclusive hadron and electron channels. In addition some of the earliest, and to date best, accelerator limits on electron muon neutrino oscillations are from Brookhaven experiments. The Laboratory is also the 'B' in the IMB underground experiment, built to search for proton decay and which caught neutrinos from the SN 1987a supernova. At present Brookhaven is heavily involved in the Gallex project in the Gran Sasso and recently a new collaboration has received scientific approval for a long baseline experiment to search for muon neutrino oscillations via muon neutrino disappearance

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

    International Nuclear Information System (INIS)

    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

  13. Operating manual for the High Flux Isotope Reactor: operating procedures

    International Nuclear Information System (INIS)

    Procedures are presented for reactor operation; instrumentation and control; reactor components; research facilities; cooling systems; containment heating, venting, and air conditioning; emergency procedures; waste systems; on-site utilities; records and data accumulation; auxiliary equipment; and technical specification requirements

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

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    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.

  15. Brookhaven highlights, October 1978-September 1979

    International Nuclear Information System (INIS)

    These highlights present an overview of the major research and development achievements at Brookhaven National Laboratory from October 1978 to September 1979. Specific areas covered include: accelerator and high energy physics programs; high energy physics research; the AGS and improvements to the AGS; neutral beam development; heavy ion fusion; superconducting power cables; ISABELLE storage rings; the BNL Tandem accelerator; heavy ion experiments at the Tandem; the High Flux Beam Reactor; medium energy physics; nuclear theory; atomic and applied physics; solid state physics; neutron scattering studies; x-ray scattering studies; solid state theory; defects and disorder in solids; surface physics; the National Synchrotron Light Source ; Chemistry Department; Biology Department; Medical Department; energy sciences; environmental sciences; energy technology programs; National Center for Analysis of Energy Systems; advanced reactor systems; nuclear safety; National Nuclear Data Center; nuclear materials safeguards; Applied Mathematics Department; and support activities

  16. High-flux beam source for cold, slow atoms or molecules

    OpenAIRE

    Maxwell, S. E.; Brahms, N.; deCarvalho, R.; Helton, J.; Nguyen, S V; Patterson, D; Doyle, J. M.; Glenn, D. R.; Petricka, J.; DeMille, D.

    2005-01-01

    We demonstrate and characterize a high-flux beam source for cold, slow atoms or molecules. The desired species is vaporized using laser ablation, then cooled by thermalization in a cryogenic cell of buffer gas. The beam is formed by particles exiting a hole in the buffer gas cell. We characterize the properties of the beam (flux, forward velocity, temperature) for both an atom (Na) and a molecule (PbO) under varying buffer gas density, and discuss conditions for optimizing these beam paramete...

  17. Development of High Flux Isotope Reactor (HFIR) subcriticality monitoring methods

    International Nuclear Information System (INIS)

    Use of subcritical source multiplication measurements during refueling has been investigated as a possible replacement for out-of-reactor subcriticality measurements formerly made on fresh HFIR fuel elements at the ORNL Critical Experiment Facility. These measurements have been used in the past for preparation of estimated critical rod positions, and as a partial verification, prior to reactor startup, that the requirements for operational shutdown margin would be met. Results of subcritical count rate data collection during recent HFIR refuelings and supporting calculations are described illustrating the intended measurement method and its expected uncertainty. These results are compared to historical uses of the out-of-reactor core measurements and their accuracy requirements, and a planned in-reactor test is described which will establish the sensitivity of the method and calibrate it for future routine use during HFIR refueling. 2 refs., 1 fig., 2 tabs

  18. High flux and high resolution VUV beam line for synchrotron radiation

    International Nuclear Information System (INIS)

    A beam line has been optimized for high flux and high resolution in the wavelength range from 30 nm to 300 nm. Sample chambers for luminescence spectroscopy on gaseous, liquid and solid samples and for photoelectron spectroscopy have been integrated. The synchrotron radiation from the storage ring DORIS (at DESY, Hamburg) emitted into 50 mrad in horizontal and into 2.2 mrad in vertical direction is focused by a cylindrical and a plane elliptical mirror into the entrance slit of a 2m normal incidence monochromator. The light flux from the exit slit is focused by a rotational elliptic mirror onto the sample yielding a size of the light spot of 4 x 0.15 mm2. The light flux at the sample reaches 7 x 1012 photons nm-1s-1 at 8 eV photon energy for a current of 100 mA in DORIS. A resolution of 0.007 nm has been obtained. (orig.)

  19. COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

    Directory of Open Access Journals (Sweden)

    Phani Kumar Domalapally

    2015-04-01

    Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

  20. A neutronic feasibility study for LEU conversion of the Brookhaven Medical Research Reactor (BMRR)

    International Nuclear Information System (INIS)

    A neutronic feasibility study for converting the Brookhaven Medical Research Reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with Brookhaven National Laboratory. Two possible LEU cores were identified that would provide nearly the same neutron flux and spectrum as the present HEU core at irradiation facilities that are used for Boron Neutron Capture Therapy and for animal research. One core has 17 and the other has 18 LEU MTR-type fuel assemblies with uranium densities of 2.5g U/cm3 or less in the fuel meat. This LEU fuel is fully-qualified for routine use. Thermal hydraulics and safety analyses need to be performed to complete the feasibility study. (author)

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

    International Nuclear Information System (INIS)

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

  2. Fusion reactor development using high power particle beams

    International Nuclear Information System (INIS)

    The present paper outlines major applications of the ion source/accelerator to fusion research and also addresses the present status and future plans for accelerator development. Applications of ion sources/accelerators for fusion research are discussed first, focusing on plasma heating, plasma current drive, plasma current profile control, and plasma diagnostics. The present status and future plan of ion sources/accelerators development are then described focusing on the features of existing and future tokamak equipment. Positive-ion-based NBI systems of 100 keV class have contributed to obtaining high temperature plasmas whose parameters are close to the fusion break-even condition. For the next tokamak fusion devices, a MeV class high power neutral beam injector, which will be used to obtain a steady state burning plasma, is considered to become the primary heating and current drive system. Development of such a system is a key to realize nuclear fusion reactor. It will be entirely indebted to the development of a MeV class high current negative deuterium ion source/accelerator. (N.K.)

  3. 1980 Annual status report: operation of the high flux reactor

    International Nuclear Information System (INIS)

    HFR Petten has been operated in 1980 in fulfilment of the 1980/83 JRC Programme Decision. Both reactor operation and utilization data have been met within a few percent of the goals set out in the annual working schedule, in support of a large variety of research programmes. Major improvements to experimental facilities have been introduced during the year and future modernization has been prepared

  4. The response time analysis of high log neutron flux rate for heavy water reactors

    International Nuclear Information System (INIS)

    The heavy water reactor such as Wolssung no. 1 has a protection/safety system named special safety system. The system has four safety systems ; shutdown no. 1, shutdown no. 2, emergency core cooling system and containment system. In this paper, the response time of high log neutron flux rate, one of the reactor trip loops of shutdown no.1/no.2, was analysed based on the description of final safety analysis report and compared to the plant measurement

  5. Progress on the realization of a new GEM based neutron diagnostic concept for high flux neutron beams

    Science.gov (United States)

    Croci, G.; Rebai, M.; Cazzaniga, C.; Palma, M. Dalla; Grosso, G.; Muraro, A.; Murtas, F.; Claps, G.; Pasqualotto, R.; Cippo, E. Perelli; Tardocchi, M.; Tollin, M.; Cavenago, M.; Gorini, G.

    2014-08-01

    Fusion reactors will need high flux neutron detectors to diagnose the deuterium-deuterium and deuterium-tritium. A candidate detection technique is the Gas Electron Multiplier (GEM). New GEM based detectors are being developed for application to a neutral deuterium beam test facility. The proposed detection system is called Close-contact Neutron Emission Surface Mapping (CNESM). The diagnostic aims at providing the map of the neutron emission due to interaction of the deuterium beam with the deuterons implanted in the beam dump surface. This is done by placing a detector in close contact, right behind the dump. CNESM uses nGEM detectors, i.e. GEM detectors equipped with a cathode that also serves as neutron-proton converter foil. After the realization and test of several small area prototypes, a full size prototype has been realized and tested with laboratory sources. Test on neutron beams are foreseen for the next months.

  6. Measurements of nuclear heating in the Petten High Flux Reactor

    International Nuclear Information System (INIS)

    The knowledge of the magnitude and distribution of the nuclear heating in all positions of the reactor core is an essential basis for the design of irradiation facilities. The principle for determining the nuclear heating is based on the caloric method. The nuclear heating is generated in a small sample suspended in the centre of a He-filled container. Since in steady state condition, the heat generation in the sample equals the heat removal over the gas annulus, the temperature difference is related to the nuclear heating in the sample. This relation is determined by means of an appropriate calibration. Two sets of three calorimeters with different sample materials, stainless steel, aluminium and graphite, are fitted in a measuring probe, which is placed in a thimble. The probe is vertically displacable in the thimble. (Auth.)

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

    International Nuclear Information System (INIS)

    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

  8. The operating experience and incident analysis for High Flux Engineering Test Reactor

    International Nuclear Information System (INIS)

    The paper describes the incidents analysis for High Flux Engineering test reactor (HFETR) and introduces operating experience. Some suggestion have been made to reduce the incidents of HFETR. It is necessary to adopt new improvements which enhance the safety and reliability of operation. (author)

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

    International Nuclear Information System (INIS)

    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

  10. Simulation of subcooled flow instability for high flux research reactors using the extended code ATHLET

    International Nuclear Information System (INIS)

    Covering the wide range of reactor safety analysis of power reactors, consisting of leak and transients, the thermohydraulic code ATHLET is being developed by the German Society for Plant and Reactor Safety (GRS). In order to extend the application range of the code to the safety analysis of low and medium flux research reactors, a model was developed and implemented permitting a description of the steam formation in the subcooled boiling regime. Considering the specific features of high flux research reactors given by both high heat flux and high flow velocity, further extension to the model of void condensation in subcooled flow has been extended and a new correlation of critical heat flux (CHF) is implemented. To validate the extended program, the thermal-hydraulic test loop (THTL) of oak ridge national laboratory (ORNL) was modeled and an extensive series of experiments concerning the onset of thermohydraulic flow instability (OFI) in subcooled boiling regime were calculated. The comparison between experiments and ATHLET post calculation shows that the extended code can accurately simulate the thermohydraulic conditions of flow instability in a wide range of heat flux up to 15 MW/m2 and inlet flow velocity up to 20 m/s. The thermohydraulic design limit characterized by the mass flux, at which the flow just becomes unstable (OFI), has been predicted in very good agreement with the experiment. However the calculated pressure drop at OFI is overestimated by a maximum deviation of about 25%. The calculated exit bulk temperature of subcooled coolant and the maximum wall temperature at OFI show a maximum deviation from experiment of 12% and 7% respectively. (author)

  11. Final report of the HFIR [High Flux Isotope Reactor] irradiation facilities improvement project

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.

  13. Simulation of subcooled flow instability for high flux research reactors using the extended code ATHLET

    International Nuclear Information System (INIS)

    Covering the wide range of reactor safety analysis of power reactors, consisting of leak and transients, the thermohydraulic code ATHLET is being developed by the Gesellschaft for Anlagen- und Reaktorsicherheit (GRS-Society for Plant and Reactor Safety) (Lerchel, G., Austregesilo, H., 1998. ATHLET Mode 1.2 Cycle A, User's Manual, GRS-p-1/Vol. 1, Rev. 1, GRS). In order to extend the code's range of application to the safety analysis of research reactors, a model was developed and implemented permitting a description of the steam formation in the subcooled boiling regime (Hainoun, A., 1994. Modellierung des unterkuehlten Siedens in ATHLET und Anwendung in wassergekuehlten Forschungsreaktoren, D 294 Diss. Univ. Bochum, Juel-2961). Considering the specific features of high flux research reactors given by both high heat flux and high flow velocity, the model of void condensation in subcooled flow has been extended and a new correlation of critical heat flux (CHF) is implemented. To validate the extended program, the Thermal-Hydraulic Test Loop (THTL) of Oak Ridge National Laboratory (ORNL) was modeled with ATHLET and an extensive series of experiments concerning the onset of thermohydraulic flow instability (OFI) in subcooled boiling regime were calculated. The comparison between experiments and ATHLET-postcalculation shows that the extended code can accurately simulate the thermohydraulic conditions of flow instability in a wide range of heat flux up to 15 MW m-2 and inlet flow velocity up to 20 m s-1. The thermohydraulic design limit characterized by the mass flux, at which the flow just becomes unstable (OFI), has been predicted in very good agreement with the experiment. However the calculated pressure drop at OFI is overestimated by a maximum deviation of about 25%. The calculated exit bulk temperature of subcooled coolant and the maximum wall temperature at OFI show a maximum deviation from experiment of 12 and 7%, respectively

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

    Indian Academy of Sciences (India)

    Usha Pal; V Jagannathan

    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 × 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.

  15. Performance Test of High Heat Flux Test Facility for the Calorimetry and Beam Control

    International Nuclear Information System (INIS)

    The Korea Heat Load Test facility, KoHLT-EB (Electron Beam) has been operating for the plasma facing components to develop fusion engineering in Korea. The ITER Neutral Beam Duct Liner (NBDL) was fabricated and tested to qualify the thermocouple fixation method for the temperature measurement during a direct collision of the high-power neutral beam during ITER operation. The NBDL is CuCrZr panels, which are actively water cooled using deep drilled channels. To perform the profile test, the assessment for the possibility of an electron beam Gaussian power density profile and the result of absorbed power for that profile before the test start is needed. To assess the possibility of Gaussian profile, for the qualification test of a Gaussian heat load profile, small calorimetry was manufactured to simulate a real heat profile in the neutral beam duct liner, and this calorimetry has two cooling channel with five thermocouples, which is the same as NBDL. Preliminary analyses with ANSYSCFX using a 3D model were performed with the calorimetry model. The heating area was modeled to be 60 mm x 250 mm. The simulated heat flux is 0.5 - 1.2 MW/m''2 at 0.75 kg/sec of the water flow rate. A steady heat flux test was performed to measure the surface heat flux, surface temperature profile. With a thermohydraulic analysis and heat load test, the Gaussian heat profile will be confirmed for this calorimetry and NBDL mockup. The Korean heat load test facility will be used to qualify the specifications of various plasma facing components in fusion devices. To conduct a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Prashant K [ORNL; Freels, James D [ORNL

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Gehin, Jess C [ORNL

    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.

  19. Approach to development of high flux research reactor with pebble-bed core

    International Nuclear Information System (INIS)

    Full text: The research nuclear reactor of a basin-type IRT with the designed power of 1 MW was put into operation in 'Sosny' settlement not far from Minsk-city in the Republic of Belarus in 1962. In 1971 after its modernization the power was increased up to 4 MW and maximum density of neutron flux in the core was: Thermal 5·1013 neutr./cm2.s Fast (E>0.8 MeV) 2·1013 neutr./cm2.s The reactor has been used for carrying out investigations in the field of solid-state physics, radiation construction materials, radiobiology, gaseous chemically reacting coolants and others. After the Chernobyl NPP accident, in the former USSR the requirements on safety of nuclear reactors have become sufficiently stricter. As to some parameters these requirements became the same as for reactors of nuclear power plants. In this connection the reactor in 'Sosny' settlement did not answer these new requirements by a number of performances such as seismicity of building, efficiency of control and protection system, corrosion in the reactor vessel and others, and it was shutdown in 1987 and its decommissioning was performed during 1988-1999. At the Joint Institute of Power and Nuclear Research -'SOSNY' have been carried out investigations on feasibility of creation of the research reactor with pebble-bed core. The concept of such reactor supposes using the following technical approaches: - Using as fuel the brought sphere micro fuel elements with the diameter of 500-750 mkm to an industrial level; - Organization of reactor operation in the regime with minimum possible fueling with 235U; - Implementation of hydraulic loading - unloading of micro fuel elements with the frequency of one or several days. Physical calculations of the core were carried out with the help of MCU-RFFI program based on the Monte-Carlo method. Two configurations of the pebble-bed core in the high flux reactor have been considered. The first configuration is the core with a neutron trap and an annular fuel layer formed

  20. Proposed fuel pin irradiation facilities for the high flux isotope reactor

    International Nuclear Information System (INIS)

    The Global Nuclear Energy Partnership (GNEP) is proposing to develop a sodium-cooled fast-spectrum reactor (SFR) to transmute and consume actinides from spent nuclear fuel. The proposed fuels include metal and oxide forms mixed actinides (U-Np-Pu-Am-Cm) as well as target concepts with perhaps both Am-Cm. The High Flux Isotope Reactor (HFIR) was built for the purpose of transmuting plutonium to various higher actinides including Am, Cm, and Cf Since a fast-spectrum irradiation facility does not exist in the United States, HFIR can fulfill a first step in the GNEP- mission that being to establish a near-term domestic capability to irradiate materials in a fast neutron spectrum. Modifications to the HFIR central target region to accomplish this goal are described. A second ongoing project for HFIR is to design capsules and installation tools and procedures to irradiate short rods of innovative nuclear fuel types and cladding materials under prototypic light water reactor (LWR) operating conditions at an accelerated rate relative to expected reactor performance. This second proposal would be for a facility representative of thermal reactor conditions rather than the GNEP concept. In order to maintain power densities within the fuel at levels normally seen by LWR reactors, an entirely new experiment and test capsule design will be needed. (authors)

  1. The high flux reactor Petten, A multi-purpose research and test facility for the future of nuclear energy

    International Nuclear Information System (INIS)

    The High Flux Reactor (HFR) at Petten, is owned by the European Commission (EC) and managed by the Institute for Advanced Materials (IAM) of the Joint Research Centre (JRC) of the EC. Its operation has been entrusted since 1962 to the Netherlands Energy Research Foundation (ECN). The HFR is one of the most powerful multi-purpose research and test reactors in the world. Together with the ECN hot cells at Petten, it has provided since three decades an integral and full complement of irradiation and examination services as required by current and future research and development for nuclear energy, industry and research organizations. Since 1963, the HFR has recognized record of consistent, reliable and high availability of more than 250 days of operation per year. The HFR has 20 in-core and 12 poolside irradiation positions, plus 12 beam tubes. With a variety of dedicated irradiation devices, and with its long-standing experience in executing small and large irradiation projects, the HFR is particularly suited for fuel, materials and components testing for all reactor lines, including thermonuclear fusion reactors. In addition, processing with neutrons and gamma rays, neutron-based research and inspection services are employed by industry and research, such as activation analysis, boron neutron capture therapy, neutron radiography and neutron diffraction. Moreover, in recent years, HFRs' mission has been broadened within the area of radioisotopes production, where, within a few years, the HFR has attained the European leadership in production volume

  2. External event probabilistic risk assessment for the High Flux Isotope Reactor (HFIR)

    International Nuclear Information System (INIS)

    The High Flux Isotope Reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988 a level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 x 10-4. In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic, wind, and fires. The overall external event contribution to core damage frequency is about 50% of the internal event initiated contribution and is dominated by seismic events

  3. Short-lived radionuclides produced on the ORNL 86-inch cyclotron and High-Flux Isotope Reactor

    International Nuclear Information System (INIS)

    The production of short-lived radionuclides at ORNL includes the preparation of target materials, irradiation on the 86-in. cyclotron and in the High Flux Isotope Reactor (HFIR), and chemical processing to recover and purify the product radionuclides. In some cases the target materials are highly enriched stable isotopes separated on the ORNL calutrons. High-purity 123I has been produced on the 86-in. cyclotron by irradiating an enriched target of 123Te in a proton beam. Research on calutron separations has led to a 123Te product with lower concentrations of 124Te and 126Te and, consequently to lower concentrations of the unwanted radionuclides, 124I and 126I, in the 123I product. The 86-in. cyclotron accelerates a beam of protons only but is unique in providing the highest available beam current of 1500 μA at 21 MeV. This beam current produces relatively large quantities of radionuclides such as 123I and 67Ga

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  5. Calculation of critical experiment parameters for the High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Six critical experiments were performed shortly before the initial ascension to power of the High Flux Isotope Reactor (HFIR). Critical configurations were determined at various control rod positions by varying the soluble boron content in the light water coolant. Calculated k-effective was 2% high at beginning-of-life (BOL) typical conditions, but was 1.0 at end-of-life (EOL) typical conditions. Axially averaged power distributions for a given radial location were frequently within experimental error. At specific r,z locations with the core, the calculated power densities were significantly different from the experimentally derived values. A reassessment of the foil activation data seems desirable

  6. Proposed Fuel Pin Irradiation Facilities for the High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    The Global Nuclear Energy Partnership (GNEP) is proposing to develop a sodium-cooled fast-spectrum reactor (SFR) to transmute and consume actinides from spent nuclear fuel. The proposed fuels include metal and oxide mixed actinides (U-Np-Pu-Am-Cm) as well as target concepts with perhaps only Am-Cm. The High Flux Isotope Reactor was built for the purpose of transmuting plutonium to various higher actinides including Am, Cm, and Cf. Since a fast-spectrum irradiation facility does not exist in the United States, HFIR can fulfill a first step in the GNEP mission; that being to establish a near-term capability to irradiate materials in a fast neutron spectrum in addition to efforts to gain access to international facilities through partnering arrangements. Modifications to the HFIR central target region to accomplish this goal are described. A second on-going project for HFIR is to design capsules and installation tools and procedures to irradiate short rods of innovative nuclear fuel types and cladding materials under prototypic LWR operating conditions at an accelerated rate relative to expected reactor performance. This second proposal would be for a facility representative of thermal reactor conditions rather than the GNEP concept. In order to maintain power densities within the fuel at levels normally seen by LWR reactors, an entirely new experiment and test capsule design will be needed than has been available in the past

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-01

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

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

    International Nuclear Information System (INIS)

    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

  9. Burn-Up Calculations for the Brookhaven Graphite Research Reactor Fuel Elements

    International Nuclear Information System (INIS)

    Fuel bum-up calculations for the Brookhaven Graphite Research Reactor involve a distribution of the thermal megawatt days of operations to the fuel elements in proportion to the average thermal neutron flux at their location in the reactor. The megawatt days so assigned can be converted to equivalent uranium-235 consumption when needed. The original fuel loading for the BGRR was neutral uranium and a single calculation was performed on each fuel element upon discharge from the reactor. A subsequent change to a fully enriched uranium-235 fuel element, however, introduced complications. The average loading of enriched uranium involves about 4800 individual elements, each occupying four different reactor positions during its term in the reactor. The total term for a central channel element is about one year as against six to eight years for an element in a peripheral channel. With the large number of individual fuel elements involved and the approximately monthly small changes needed for operation, it was necessary to resort to a computer programme to follow the burn-up of all the elements on the reactor continuously. Both this and other functions of the computer programme are discussed in the paper. To date, uranium has been recovered from two batches of spent fuel. On the first, involving 3674 elements discharged from the reactor over a period of 4.9 years, the recovery figures were 5.5% higher than the calculated total of 32.3 kg uranium-235. On the second batch, involving 1296 elements discharged from the reactor over a period of one year, the recovery figures were 2.3% higher than the calculated figures of 10.8 kg uranium-235. This relatively close agreement seems to indicate that the assumptions made to simplify the programme are acceptable and that the results of the programme are satisfactory for our particular accounting and operating requirements. (author)

  10. High flux cold Rubidium atomic beam for strongly coupled Cavity QED

    CERN Document Server

    Roy, Basudev

    2012-01-01

    This paper presents a setup capable of producing a high-flux continuous beam of cold rubidium atoms for cavity QED experiments in the regime of strong coupling. A 2 $D^+$ MOT, loaded by rubidium getters in a dry film coated vapor cell, fed a secondary moving-molasses MOT (MM-MOT) at a rate of 1.5 x $10^{10}$ atoms/sec. The MM-MOT provided a continuous beam with tunable velocity. This beam was then directed through the waist of a 280 $\\mu$m cavity resulting in a Rabi splitting of more than +/- 10 MHz. The presence of sufficient number of atoms in the cavity mode also enabled splitting in the polarization perpendicular to the input. The cavity was in the strong coupling regime, with parameters (g, $\\kappa$, $\\gamma$)/2$\\pi$ equal to (7, 3, 6)/ 2$\\pi$ MHz.

  11. Countercurrent flow limited (CCFL) heat flux in the high flux isotope reactor (HFIR) fuel element

    International Nuclear Information System (INIS)

    The countercurrent flow (CCF) performance in the fuel element region of the HFIR is examined experimentally and theoretically. The fuel element consists of two concentric annuli filled with aluminum clad fuel plates of 1.27 mm thickness separated by 1.27 mm flow channels. The plates are curved as they go radially outward to accomplish constant flow channel width and constant metal-to-coolant ratio. A full-scale HFIR fuel element mock-up is studied in an adiabatic air-water CCF experiment. A review of CCF models for narrow channels is presented along with the treatment of CCFs in system of parallel channels. The experimental results are related to the existing models and a mechanistic model for the ''annular'' CCF in a narrow channel is developed that captures the data trends well. The results of the experiment are used to calculate the CCFL heat flux of the HFIR fuel assembly. It was determined that the HFIR fuel assembly can reject 0.62 Mw of thermal power in the CCFL situation. 31 refs., 17 figs

  12. Dynamic response of the high flux isotope reactor structure caused by nearby heavy load drop

    International Nuclear Information System (INIS)

    A heavy load of 50,000 lb is assumed to drop from 10 ft above the bottom of the High Flux Isotope Reactor (HFIR) pool at the loading station. The consequences of the dynamic impact to the bottom slab of the pool and to the nearby HFIR reactor vessel are analyzed by applying the ABAQUS computer code The results show that both the BM vessel structure and its supporting legs are subjected to elastic disturbances only and, therefore, will not be damaged. The bottom slab of the pool, however, will be damaged to about half of the slab thickness. The velocity response spectrum at the concrete floor next to the HFIR vessel as a result of the vibration caused by the impact is obtained. It is concluded, that the damage caused by heavy load drop at the loading station is controlled by the slab damage and the nearby HFIR vessel and the supporting legs will not be damaged

  13. Facility modification for severe accident mitigation at the High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    As a part of the recently completed Safety Analysis Report (SAR) for the High Flux Isotope Reactor (HFIR), a detailed MELCOR model was developed for use in the analysis of selected events that lead to core damage. This detailed MELCOR model included the HFIR's reactor coolant system, dynamic confinement, building and yard piping, and a model of the outside waste collection storage tanks. Analyses of several accident sequences involving large breaks in the cold leg piping in the heat exchanger cells and pipe tunnel were conducted with the model. From the results of these simulations, a need for modification of the HFIR confinement was identified. These confinement modifications resulted in over a 12 hour delay in the release of primary system water (an any fission products it contains) outside of the HFIR confinement. This delay would provide valuable time to prevent or mitigate any releases to the environment due to a break in the primary system of the HFIR

  14. Calculations for HFIR [High Flux Isotope Reactor] fuel plate non- bonding and fuel segregation uncertainty factors

    International Nuclear Information System (INIS)

    The effects of non-bonds and of fuel segregation on the package factors of the heat flux in the High Flux Isotope Reactor (HFIR) are examined. The effects of the two defects are examined both separately and together. It is concluded that the peaking factors that are used in the present HFIR thermal analysis code are conservative and thus no changes in the peaking factors are necessary to continue to ensure that HFIR is safe. A study was made of the effect of the non-bond spot diameter on the peaking factor. The conclusion is that the spot can have diameter more than three times the maximum value allowed by the specifications before the peaking factor is greater than the maximum value specified in the present HFIR thermal analysis code. 6 refs., 7 figs., 8 tabs

  15. Laser ion source with solenoid for Brookhaven National Laboratory-electron beam ion sourcea)

    Science.gov (United States)

    Kondo, K.; Yamamoto, T.; Sekine, M.; Okamura, M.

    2012-02-01

    The electron beam ion source (EBIS) preinjector at Brookhaven National Laboratory (BNL) is a new heavy ion-preinjector for relativistic heavy ion collider (RHIC) and NASA Space Radiation Laboratory (NSRL). Laser ion source (LIS) is a primary ion source provider for the BNL-EBIS. LIS with solenoid at the plasma drift section can realize the low peak current (˜100 μA) with high charge (˜10 nC) which is the BNL-EBIS requirement. The gap between two solenoids does not cause serious plasma current decay, which helps us to make up the BNL-EBIS beamline.

  16. Laser ion source with solenoid for Brookhaven National Laboratory-electron beam ion source

    International Nuclear Information System (INIS)

    The electron beam ion source (EBIS) preinjector at Brookhaven National Laboratory (BNL) is a new heavy ion-preinjector for relativistic heavy ion collider (RHIC) and NASA Space Radiation Laboratory (NSRL). Laser ion source (LIS) is a primary ion source provider for the BNL-EBIS. LIS with solenoid at the plasma drift section can realize the low peak current (∼100 μA) with high charge (∼10 nC) which is the BNL-EBIS requirement. The gap between two solenoids does not cause serious plasma current decay, which helps us to make up the BNL-EBIS beamline.

  17. Laser ion source with solenoid for Brookhaven National Laboratory-electron beam ion source.

    Science.gov (United States)

    Kondo, K; Yamamoto, T; Sekine, M; Okamura, M

    2012-02-01

    The electron beam ion source (EBIS) preinjector at Brookhaven National Laboratory (BNL) is a new heavy ion-preinjector for relativistic heavy ion collider (RHIC) and NASA Space Radiation Laboratory (NSRL). Laser ion source (LIS) is a primary ion source provider for the BNL-EBIS. LIS with solenoid at the plasma drift section can realize the low peak current (∼100 μA) with high charge (∼10 nC) which is the BNL-EBIS requirement. The gap between two solenoids does not cause serious plasma current decay, which helps us to make up the BNL-EBIS beamline. PMID:22380298

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

    International Nuclear Information System (INIS)

    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: (1) the probability of vessel fracture due to hydrotest vs vessel life at several hydrotest pressures, (2) the hydrotest time interval vs the uncertainty of the nil ductility temperature increase rate, and (3) 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

  19. Estimation of HFIR [High-Flux Isotope Reactor] core flow rate using a back propagation network

    International Nuclear Information System (INIS)

    The corrosion of aluminum dispersion fuel of Oak Ridge National Laboratory's High-Flux Isotope Reactor (HFIR) has recently been under review. The surface temperature of the fuel under high-flux conditions increases as the oxide layer grows or crud deposits on the fuel. This additional temperature rise in turn speeds the oxide buildup creating an unstable situation. Since the low thermal conductivity of the corrosion product film may compromise the thermal limits of the reactor, and also, erosion of the aluminum plates may lead to leakage of fission products from the plates, a careful monitoring of corrosion would be useful for the HFIR. In addition, an examination of HFIR operating data indicates that excessive corrosion has a noticeable effect on core flow rate and pressure drop. Typically, as the thickness of oxide layer increases, the flow rate gradually decreases and the pressure drop increases; therefore, it has been recommended that the core pressure drop and flow characteristics be monitored during each fuel cycle. The purpose of this study is to investigate the feasibility of applying neural networks to predict the flow rate in the HFIR core using a set of other HFIR operating data. The back propagation network (BPN) is a multilayer, fully connected heteroassociative network. During the training stage, the BPN learning algorithm computes the weights between pairs such that the difference between the actual output and the network output is minimized in a least-squares sense

  20. Development of a high-heat-flux target for multimegawatt, multisecond neutral beams at ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Combs, S.K.; Milora, S.L.; Bush, C.E.; Foster, C.A.; Haselton, H.H.; Hayes, P.H.; Menon, M.M.; Moeller, J.A.; Sluss, F.; Tsai, C.C.

    1984-01-01

    A high-heat-flux target has been developed for intercepting multimegawatt, multisecond neutral beam power at the Oak Ridge National Laboratory (ORNL). Water-cooled copper swirl tubes are used for the heat transfer medium; these tubes exhibit an enhancement in burnout heat flux over conventional axial-flow tubes. The target consists of 126 swirl tubes (each 0.95 cm in outside diameter with 0.16-cm-thick walls and approx. =1 m long) arranged in a V-shape. Two arrays of parallel tubes inclined at an angle ..cap alpha.. to the beam axis form the V-shape, and this geometry reduces the surface heat flux by a factor of 1/sin ..cap alpha.. (for the present design, ..cap alpha.. =13/sup 0/ and 21/sup 0/). In tests with the ORNL long-pulse ion source (13- by 43-cm grid), the target has handled up to 3-MW, 30-s beam pulses with no deleterious effects. The peak power density was estimated at approx. =15 kW/cm/sup 2/ normal to the beam axis (5.4 kW/cm/sup 2/ maximum on tube surfaces). The water flow rate through the target was 41.6 L/s (660 gpm) or 0.33 L/s (5.2 gpm) per tube (axial flow velocity = 11.6 m/s). The corresponding pressure drop across the target was 1.14 MPa (165 psi) with an inlet pressure of 1.45 MPa (210 psia). Data are also presented from backup experiments in which individual tubes were heated by a small ion source (10-cm-diam grid) to characterize tube performance. These results suggest that the target should handle peak power densities in the range 25 to 30 kW/cm/sup 2/ normal to the beam axis (approx. =10 kW/cm/sup 2/ maximum on tube surfaces) with the present flow parameters. This translates to beam power levels of 5 to 6 MW for equivalent beam optics.

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

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD)

    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.

  2. Loading beryllium targets to extend the high flux isotope reactor's cycle length

    International Nuclear Information System (INIS)

    Various arrangements of beryllium loadings to create an internal neutron reflector in the flux trap region of the Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR) have been investigated. In particular, the impact upon fuel cycle length has been studied by performing calculations using the HFIR MCNP-based model HFV4.0. This study included examining perturbations in reactivity, flux, and power distribution caused by the various beryllium loadings. The HFIR Cycle 400 core configuration was used as a reference to calculate the impact of beryllium loadings upon cycle length. Three different configurations of beryllium loadings were investigated and compared against the Cycle 400 benchmark calculations; Cases 1 through 3 modeled combinations of 12 and 18 beryllium rods loaded into unused experimental sites. Calculated eigenvalues have shown that potential increases in reactivity between 0.56 and 0.79 dollars are attainable, depending on the various beryllium configurations. These results correspond to possible increases in fuel cycle length ranging between 2.3% and 3.3%. On the basis of their practicality, cost versus benefit, and greater potential for implementation, Cases 2 and 3 (both with 18 beryllium rods) were studied further and are herein reported in greater detail. Neutron flux distributions for Cases 2 and 3 were calculated at the horizontal mid-plane of the flux trap region, which showed no significant changes in the thermal flux magnitude and radial profile in comparison to Cycle 400. Likewise, safety analysis related parameters were contrasted, revealing power increments of up to 2% near the inside edge of the inner fuel element, well below the maximum acceptable value of 9%, a standing guideline employed for experiments at the HFIR. Additionally, the average neutron heat generation rate in beryllium rods and the maximum heat generation rate were evaluated to confirm that the design provides adequate coolant flow inside the rod and around the

  3. Total quality management for addressing suspect parts at the Oak Ridge High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Martin Marietta Energy System (MMES) Research Reactors Division (RRD), operator of the High Flux Isotope Reactor (HFIR) recently embarked on an aggressive Program to address the issue of suspect Parts and to enhance their procurement process. Through the application of TQM process improvement, RRD has already achieved improved efficiency in specifying, procuring, and accepting replacement items for its largest research reactor. These process improvements have significantly decreased the risk of installing suspect parts in the HFIR safety systems. To date, a systematic plan has been implemented, which includes the following elements: Process assessment and procedure review; Procedural enhancements; On-site training and technology transfer; Enhanced receiving inspections; Performance supplier evaluations and source verifications integrated processes for utilizing commercial grade products in nuclear safety-related applications. This paper will describe the above elements, how a partnership between MMES and Gilbert/Commonwealth facilitated the execution of the plan, and how process enhancements were applied. We will also present measures for improved efficiency and productivity, that MMES intends to continually address with Quality Action Teams

  4. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE BROOKHAVEN GRAPHITE RESEARCH REACTOR ENGINEERED CAP, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK DCN 5098-SR-07-0

    Energy Technology Data Exchange (ETDEWEB)

    Evan Harpenau

    2011-07-15

    The Oak Ridge Institute for Science and Education (ORISE) has reviewed the project documentation and data for the Brookhaven Graphite Research Reactor (BGRR) Engineered Cap at Brookhaven National Laboratory (BNL) in Upton, New York. The Brookhaven Science Associates (BSA) have completed removal of affected soils and performed as-left surveys by BSA associated with the BGRR Engineered Cap. Sample results have been submitted, as required, to demonstrate that remediation efforts comply with the cleanup goal of {approx}15 mrem/yr above background to a resident in 50 years (BNL 2011a).

  5. Calculational study on irradiation of americium fuel samples in the Petten High Flux Reactor

    International Nuclear Information System (INIS)

    A calculational study on the irradiation of americium samples in the Petten High Flux Reactor (HFR) has been performed. This has been done in the framework of the international EFTTRA cooperation. For several reasons the americium in the samples is supposed to be diluted with a neutron inert matrix, but the main reason is to limit the power density in the sample. The low americium nuclide density in the sample (10 weight % americium oxide) leads to a low radial dependence of the burnup. Three different calculational methods have been used to calculate the burnup in the americium sample: Two-dimensional calculations with WIMS-6, one-dimensional calculations with WIMS-6, and one-dimensional calculations with SCALE. The results of the different methods agree fairly well. It is concluded that the radiotoxicity of the americium sample can be reduced upon irradiation in our scenario. This is especially the case for the radiotoxicity between 100 and 1000 years after storage. (orig.)

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

    International Nuclear Information System (INIS)

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

  7. Efforts to Reduce Radioactive Wastes at High Flux Advanced Neutron Application Reactor (HANARO) in Korea

    International Nuclear Information System (INIS)

    The High-flux Advanced Neutron Application Reactor (HANARO) has the equipment to treat the gaseous radioactive waste generated within itself but it does not have proper ways to treat the waste in either a liquid or solid form. For the last 5 years, every effort has been made to reduce the radioactive wastes in HANARO. Improvements of the equipment and operating procedures regarding the generation of radioactive waste in the field have resulted in an effective reduction of the radioactive waste. In addition, as an outgrowth of the research efforts in connection with the demand in the field, several new methods have been developed to effectively treat the radioactive waste. Efforts to reduce and treat the radioactive waste in HANARO will continue and that will contribute greatly to an improvement of the reliability and safety of HANARO. (authors)

  8. Aging, maintenance and modernization of instrumentation at the ORNL High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    This paper describes actions taken at ORNL to upgrade and modernize systems associated with the High Flux Isotope Reactor (HFIR). Three systems are described. The first is a redesigned resistance temperature device commonly used as a temperature detector at HFIR. The aging of existing devices, and lack of spare devices prompted the redesign of new temperature sensors which used commercial grade sensors in a redesigned assembly which allows easier maintenance. The second is a newly designed neutron detector system to replace existing aging devices. The new design uses commercial ionization chambers, in a cheaper, simpler, and less complicated design which could fit in the previous space, and provide monitoring for control and for protection. The third is the development and implementation of a new test procedure for checking the safety performance of the magnetic safety rod release system. This new procedure allows for electronic testing of the modules with considerably lessened chances that a rod will be dropped during the weekly testing

  9. High energy physics at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    The high energy plans at BNL are centered around the AGS and ISABELLE, or a variant thereof. At present the AGS is maintaining a strong and varied program. This last year a total of 4 x 1019 protons were delivered on target in a period of approximately 20 weeks. Physics interest is very strong, half of the submitted proposals are rejected (thereby maintaining high quality experiments) and the program is full over the next two years. The future colliding beam facility will utilize the AGS as an injector and will be a dedicated facility. It will have six intersection regions, run > 107 sec/year, and explore a new domain of energy and luminosity. Common to all the considered alternatives is a large aperture proton ring. These possible choices involve pp, ep, and heavy ion variants. The long term philosophy is to run the AGS as much as possible, continuously to upgrade it in performance and reliability, and then to phase it down as the new collider begins operation

  10. Elaboration of mini plates with U-Mo for irradiation in a high flux reactor

    International Nuclear Information System (INIS)

    Full text: International new efforts for the reconversion of HEU in research, testing and radioisotopes production reactors, have greatly incremented U-Mo fuels qualification activities. These qualifications require the resolution of undesired interaction at high fluxes between UMo particles and the aluminum matrix in the case of dispersed fuels and the development of U-Mo monolithic fuels. These efforts are being manifested in the planning and execution of additional series of irradiation tests of mini plates and full size plates. Recently, CNEA has elaborated mini plates with different proposals for the irradiation at the ATR reactor (250 MWTH, maximum thermal neutron flux 1015 n.cm-2.seg-1) at Idaho National Laboratory, USA. Uranium 7% (w/w) molybdenum (U-7Mo) particles were coated with silicon. Chemical vapour deposition (CVD) of silane and high temperature diffusion of silicon were used. Hydrided, milled and dehydrated (HMD) particles heat treated at 1000 C degrees during four hours and centrifugal atomized powder were coated and the results compared. Mini plates were elaborated with both kinds of particles. Mini plates were also elaborated with U-7Mo and silicon particles dispersed in the aluminium matrix. Monolithic mini plates were also developed by co lamination of U-7Mo with a Zircaloy-4 cladding. The different steps of this process are detailed and the method is shown to be versatile, can be easily scaled up and is performed with small modifications of usual equipment in fuel plants. The irradiation experiment is called RERTR-7A, includes a total of 32 mini plates and it is planed to finalize by mid 2006. (author)

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

    International Nuclear Information System (INIS)

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

  12. Surface modification of molten W exposed to high heat flux helium neutral beams

    International Nuclear Information System (INIS)

    High heat flux tests with central heat flux of 10.5 MW/m2 using helium neutral beams have been carried out on rolled tungsten. The energy of helium particles is 33 keV and the particle flux is 2 × 1021 m−2 s−1. An 80 × 65 × 3 mm3 rolled tungsten plate is firstly exposed to a 4.6 s pulse resulting in partially molten surfaces. Thereafter the tungsten plate is irradiated by several helium pulses with fluences of 1.2–2.5 × 1022/m2 and peak temperatures from 1450 to 2590 °C. The experiments show that: (1) helium-induced surface modification of the resolidified tungsten surface is very different from that of the non-molten surface; (2) the surface morphology of molten surface is closely related to the orientation of the resolidified grain; (3) the evolution of surface modifications, for both of the molten and non-molten tungsten surfaces, indicates a strong dependence on the surface temperature and local helium fluence

  13. Brookhaven highlights, October 1978-September 1979. [October 1978 to September 1979

    Energy Technology Data Exchange (ETDEWEB)

    1979-01-01

    These highlights present an overview of the major research and development achievements at Brookhaven National Laboratory from October 1978 to September 1979. Specific areas covered include: accelerator and high energy physics programs; high energy physics research; the AGS and improvements to the AGS; neutral beam development; heavy ion fusion; superconducting power cables; ISABELLE storage rings; the BNL Tandem accelerator; heavy ion experiments at the Tandem; the High Flux Beam Reactor; medium energy physics; nuclear theory; atomic and applied physics; solid state physics; neutron scattering studies; x-ray scattering studies; solid state theory; defects and disorder in solids; surface physics; the National Synchrotron Light Source ; Chemistry Department; Biology Department; Medical Department; energy sciences; environmental sciences; energy technology programs; National Center for Analysis of Energy Systems; advanced reactor systems; nuclear safety; National Nuclear Data Center; nuclear materials safeguards; Applied Mathematics Department; and support activities. (GHT)

  14. Water flow characteristics of Baumkuchen type fuel elements for Kyoto University high neutron flux reactor

    International Nuclear Information System (INIS)

    The Kyoto University high neutron flux reactor is a light water-moderated and cooled, divided core type reactor with heavy water reflector. In the core, six inside fuel elements and twelve outside fuel elements are arranged in double ring form, and two cylindrical, divided cores are placed at 15 cm distance. The flow rate distribution and pressure loss in the fuel elements constitute the base of the thermo-hydraulic design of the core, therefore the model fuel elements of full size were made, and the water flow experiment was carried out to examine their characteristics. It was found that the flow velocity in channels was strongly affected by the accuracy of channel gaps. The calculation of pressure loss in fuel elements, the experiments on inside fuel elements and outside fuel elements, and the results of experiments such as the calibration of the cooling channels in outside fuel elements, the relation between total flow rate and pressure loss, and the characteristics of flow at the time of reverse flow are reported. The general characteristics of flow in fuel elements were in good agreement with the prediction. In the pressure loss in fuel elements, the friction between fuel plates and the resistance of nozzles were the controlling factors under the rated operating conditions of the HFR. (Kako, I.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  16. High-flux cold rubidium atomic beam for strongly-coupled cavity QED

    International Nuclear Information System (INIS)

    This paper presents a setup capable of producing a high-flux continuous beam of cold rubidium atoms for cavity quantum electrodynamics experiments in the region of strong coupling. A 2D+ magneto-optical trap (MOT), loaded with rubidium getters in a dry-film-coated vapor cell, fed a secondary moving-molasses MOT (MM-MOT) at a rate greater than 2 x 1010 atoms/s. The MM-MOT provided a continuous beam with a tunable velocity. This beam was then directed through the waist of a cavity with a length of 280 μm, resulting in a vacuum Rabi splitting of more than ±10 MHz. The presence of a sufficient number of atoms in the cavity mode also enabled splitting in the polarization perpendicular to the input. The cavity was in the strong coupling region, with an atom-photon dipole coupling coefficient g of 7 MHz, a cavity mode decay rate κ of 3 MHz, and a spontaneous emission decay rate γ of 6 MHz.

  17. High-flux cold rubidium atomic beam for strongly-coupled cavity QED

    Energy Technology Data Exchange (ETDEWEB)

    Roy, Basudev [Indian Institute of Science Education and Research, Kolkata (India); University of Maryland, MD (United States); Scholten, Michael [University of Maryland, MD (United States)

    2012-08-15

    This paper presents a setup capable of producing a high-flux continuous beam of cold rubidium atoms for cavity quantum electrodynamics experiments in the region of strong coupling. A 2D{sup +} magneto-optical trap (MOT), loaded with rubidium getters in a dry-film-coated vapor cell, fed a secondary moving-molasses MOT (MM-MOT) at a rate greater than 2 x 10{sup 10} atoms/s. The MM-MOT provided a continuous beam with a tunable velocity. This beam was then directed through the waist of a cavity with a length of 280 μm, resulting in a vacuum Rabi splitting of more than ±10 MHz. The presence of a sufficient number of atoms in the cavity mode also enabled splitting in the polarization perpendicular to the input. The cavity was in the strong coupling region, with an atom-photon dipole coupling coefficient g of 7 MHz, a cavity mode decay rate κ of 3 MHz, and a spontaneous emission decay rate γ of 6 MHz.

  18. Application of expert systems to heat exchanger control at the 100-megawatt high-flux isotope reactor

    International Nuclear Information System (INIS)

    The High-Flux Isotope Reactor (HFIR) is a 100-MW pressurized water reactor at the Oak Ridge National Laboratory. It is used to produce isotopes and as a source of high neutron flux for research. Three heat exchangers are used to remove heat from the reactor to the cooling towers. A fourth heat exchanger is available as a spare in case one of the operating heat exchangers malfunctions. It is desirable to maintain the reactor at full power while replacing the failed heat exchanger with the spare. The existing procedures used by the operators form the initial knowledge base for design of an expert system to perform the switchover. To verify performance of the expert system, a dynamic simulation of the system was developed in the MACLISP programming language. 2 refs., 3 figs

  19. A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor. Vol 1

    International Nuclear Information System (INIS)

    The Department of Industry, Science and Tourism selected PLG, an EQE International Company, to systematically and independently evaluate the safety of the High Flux Australian Reactor (HIFAR), located at Lucas Heights, New South Wales. PLG performed a comprehensive probabilistic safety assessment (PSA) to quantify the risks posed by operation of HIFAR . The PSA identified possible accident scenarios, estimated their likelihood of occurrence, and assigned each scenario to a consequence category; i.e., end state. The accident scenarios developed included the possible release of radioactive material from irradiated nuclear fuel and of tritium releases from reactor coolant. The study team developed a recommended set of safety criteria against which the results of the PSA may be judged. HIFAR was found to exceed one of the two primary safety objectives and two of the five secondary safety objectives. Reactor coolant leaks, earthquakes, and coolant pump trips were the accident initiators that contributed most to scenarios that could result in fuel overheating. Scenarios initiated by earthquakes were the reason the frequency criterion for the one primary safety objective was exceeded. Overall, the plant safety status has been shown to be generally good with no evidence of major safety-related problems from its operation. One design deficiency associated with the emergency core cooling system was identified that should be corrected as soon as possible. Additionally, several analytical issues have been identified that should be investigated further. The results from these additional investigations should be used to determine whether additional plant and procedural changes are required, or if further evaluations of postulated severe accidents are warranted. Supporting information can be found in Appendix A for the seismic analysis and in the Appendix B for selected other external events

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

    International Nuclear Information System (INIS)

    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 RTNDT 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 10-4. 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

  1. A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor. Vol 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    The Department of Industry, Science and Tourism selected PLG, an EQE International Company, to systematically and independently evaluate the safety of the High Flux Australian Reactor (HIFAR), located at Lucas Heights, New South Wales. PLG performed a comprehensive probabilistic safety assessment (PSA) to quantify the risks posed by operation of HIFAR . The PSA identified possible accident scenarios, estimated their likelihood of occurrence, and assigned each scenario to a consequence category; i.e., end state. The accident scenarios developed included the possible release of radioactive material from irradiated nuclear fuel and of tritium releases from reactor coolant. The study team developed a recommended set of safety criteria against which the results of the PSA may be judged. HIFAR was found to exceed one of the two primary safety objectives and two of the five secondary safety objectives. Reactor coolant leaks, earthquakes, and coolant pump trips were the accident initiators that contributed most to scenarios that could result in fuel overheating. Scenarios initiated by earthquakes were the reason the frequency criterion for the one primary safety objective was exceeded. Overall, the plant safety status has been shown to be generally good with no evidence of major safety-related problems from its operation. One design deficiency associated with the emergency core cooling system was identified that should be corrected as soon as possible. Additionally, several analytical issues have been identified that should be investigated further. The results from these additional investigations should be used to determine whether additional plant and procedural changes are required, or if further evaluations of postulated severe accidents are warranted. Supporting information can be found in Appendix A for the seismic analysis and in the Appendix B for selected other external events refs., 139 tabs., 85 figs. Prepared for Department of Industry, Science and Tourism

  2. Irradiation effects in fused quartz 'Suprasil' as a detector of fission fragments under high flux of reactor neutrons

    International Nuclear Information System (INIS)

    A systematic study about the registration characteristics of synthetic fused quartz 'Suprasil I' use as a detector of fission fragments under high flux of reactor neutrons and the effects of irradiation on it was performed. Fission fragments of 252Cf, gamma radiation doses of of 60Co up to 150 MGy, and integrated neutrons fluxes up to 1020 n/cm2 were used. A model to explain the effects on track registration and development characteristics of 'Suprasil I' irradiated on reactors were proposed, based on the obtained results for efficiency an for annealing. (C.G.C.)

  3. High heat flux testing of B4C/Cu and SiC/Cu functionally graded materials simulated by laser and electron beam

    International Nuclear Information System (INIS)

    B4C, SiC and C, Cu functionally graded-materials (FGMs) have been developed by plasma spraying and hot pressing. Their high-heat flux properties have been investigated by high energy laser and electron beam for the simulation of plasma disruption process of the future fusion reactors. And a study on eroded products of B4C/Cu FGM under transient thermal load of electron beam was performed. In the experiment, SEM and EDS analysis indicated that B4C and SiC were decomposed, carbon was preferentially evaporated under high thermal load, and a part of Si and Cu were melted, in addition, the splash of melted metal and the particle emission of brittle destruction were also found. Different erosive behaviors of carbon-based materials (CBMs) caused by laser and electron beam were also discussed

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Guida, Tracey [University of Pittsburgh

    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.

  6. Job/task analysis for I ampersand C [Instrumentation and Controls] instrument technicians at the High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR); Chandler, David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR); Ade, Brian J [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR); Sunny, Eva E [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR); Betzler, Benjamin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR); Pinkston, Daniel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). High Flux Isotope Reactor (HFIR)

    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.

  9. Temperature dependence of swelling in Type 316 stainless steel irradiated in HFIR [High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    The temperature dependence of swelling was investigated in solution-annealed (SA) and 20% cold-worked (CW) type 316 stainless steel irradiated to 30 dpa at 300 to 6000C in the High Flux Isotope Reactor (HFIR). At irradiation temperatures ≤ 4000C, a high concentration (2 to 4 x 1023 m-3) of small bubbles (1.5 to 4.5 nm diam) formed uniformly in the matrix. Swelling was low (0C. At 5000C, there was a mixture of bubbles and voids, but at 6000C, most of the cavities were voids. Maximum swelling (∼5%) occurred at 5000C. By contrast, cavities in 20% CW specimens were much smaller, with diameters of 6 and 9 nm at 500 and 6000C, respectively, suggesting that they were primarily bubbles. The cavity number density in the CW 316 at both 500 and 6000C (∼1 x 1022 m-3) was about one order of magnitude less than at 4000C. Swelling increased slightly as irradiation temperature increased, peaking at 6000C (0.3%). These results indicate that SA 316 swells more than CW 316 at 500 and 6000C, but both SA and CW 316 are resistant to void swelling in HFIR at 4000C and below to 30 dpa. 15 refs

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

    Energy Technology Data Exchange (ETDEWEB)

    Kohring, M.W.

    1987-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Travis, Adam R [ORNL

    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.

  13. A Preliminary Calculation of Annular Core Design for a High-flux Advanced Research Reactor

    International Nuclear Information System (INIS)

    Many of research reactors in operation over the world become old and the number of research reactors is expected to be reduced around 1/3 within a next decade. So it may be necessary to prepare in advance for the future demands of research reactors with a high performance. Therefore, based on the HANARO experiences through design to operation, a concept development of an improved research reactor is under doing. In this paper, 10 MW conceptual annular core is proposed and its basic characteristics were analyzed as a preliminary step

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

    International Nuclear Information System (INIS)

    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 β-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

  15. Postirradiation properties of the 6061-T6 aluminum High Flux Isotope Reactor hydraulic tube

    International Nuclear Information System (INIS)

    A tube of 6061 aluminum alloy in a T6 temper, precipitation-hardened with Mg2Si, was examined after irradiation in the core of the High Flux Isotope Reactor to fluences up to 1.3 x 1023 neutrons (n)/cm2 (0.1 MeV) and 3.1 x 1023 n/cm2 (thermal) in contact with the cooling water at a temperature of about 550C. The alloy displayed up to 2.5 percent swelling due mainly to a precipitate of transmutation-produced silicon of which more than 6 weight percent was formed. Some cavities were also observed. Tension tests in the temperature range 55 to 2000C showed radiation-induced increases in yield stresses and ultimate stresses of 50 to 80 percent; elongation was reduced from the range 10 to 15 percent to about 5 percent at 550C and to about 3 percent at test temperatures above 1000C. The fracture mode was changed from transgranular tearing around inclusions to a mixture of transgranular tearing and ductile intergranular separation. These changes are attributed primarily to the radiation-induced silicon precipitate. A rim of intergranular cracks formed at the originally oxidized surfaces of the tube duringtension testing and became deeper with increasing neutron irradiation and increasing temperature

  16. Postirradiation properties of the 6061-T6 aluminum high flux isotope reactor hydraulic tube

    International Nuclear Information System (INIS)

    A tube of 6061 aluminum alloy in a T6 temper, precipitation-hardened with Mg2Si, was examined after irradiation in the core of the High Flux Isotope Reactor to fluences up to 1.3 x 1023 neutrons (n)/cm2 (0.1 MeV) and 3.1 x 1023 n/cm2 (thermal) in contact with the cooling water at a temperature of about 550C. The alloy displayed up to 2.5 percent swelling due mainly to a precipitate of transmutaion-produced silicon of which more than 6 weight perent was formed. Some cavities were also observed. Tension tests in the temperature range 55 to 2000C showed radiation-induced increases in yield stresses and ultimate stresses of 50 to 80 percent; elongation was reduced from the range 10 to 15 percent to about 5 percent at 55C0 and to about 3 percent at test temperatures above 100C0. The fracture mode was changed from transgranular tearing around inclusions to a mixture of transgranular tearing and ductile intergradular separation. These changes are attributed primarily to the radiation-induced silicon precipitate. A rim of intergranular cracks formed at the originally oxidized surfaces of the tube during tension testing and became deeper with increasing neutron irradiation and increasing temperature

  17. The survey on the supporting ground on the construction site of High Flux Reactor Building in Research Reactor Institute of Kyoto University

    International Nuclear Information System (INIS)

    As part of the seismic design of the High Flux Reactor building which is planned to be constructed by Kyoto University Research Reactor Institute, the stability of the supporting ground has been analyzed. This report concerns the ground survey which has been carried out to obtain the basic data on the supporting ground. The outline of the ground around the construction site of High Flux Reactor has been already made clear by the last survey. Therefore, the purpose of this ground survey is mainly to make clear the mechanical properties of the soil. The survey has been carried out concerning the supporting ground and several layers deeper than that. The main items obtained are as follows. (1) modulus of deformation (2) breaking strength and creep strength (3) coefficient of permeability (4) ground water level. (author)

  18. Study of thermohydraulic instability and design limits of high flux research reactors using the extended code ATHLET

    International Nuclear Information System (INIS)

    Covering the wide range of reactor safety analysis of power reactors, consisting of leak and transients, the thermohydraulic code ATHLET is being developed by the German Society for Plant and Reactor Safety (GRS). In order to extend the application range of the code to the safety analysis of low and medium flux research reactors, a model was developed and implemented permitting a description of the steam formation in the subcooled boiling regime. Considering the specific features of high flux research reactors given by both high heat flux and high flow velocity, further extension to the model of void condensation in subcooled flow has been extended and a new correlation of critical heat flux (CHF) is implemented. To validate the extended Program, the Thermal Hydraulic Test Loop (THTL) of Oak ridge National Laboratory (ORNL) was modeled and an extensive series of experiments concerning the onset of thermohydraulic flow instability (OFI) in subcooled boiling regime were calculated. The comparison between experiments and ATHLET-postcalculation shows that the extended code can accurately simulate the thermohydraulic conditions of flow instability in a wide range of heat flux up to 15 MW/m2 and inlet flow velocity up to 20 m/s. The thermohydraulic design limit characterized by the mass flux, at which the flow just becomes unstable (OFI), has been predicted in very good agreement with the experiment. However the calculated pressure drop at OFI is overestimated by a maximum deviation of about 25%. The calculated exit bulk temperature of subcooled coolant and the maximum wall temperature at OFI show a maximum deviation from experiment of 12% and 7% respectively. The extended code has been applied successfully to simulate the flow reversal in the fuel element of German high flux research reactor FRM-II. This phenomenon is expected in case of shutdown pumps failure. The results show the code's capability to simulate the flow reversal from down ward to up ward direction

  19. Rf beam loading in the Brookhaven AGS with booster injection

    International Nuclear Information System (INIS)

    Multi-batch bunched beam loading during injection from the Booster to the AGS will be discussed. The full intensity beam injection to the upgraded AGS rf system with beam phase and radial feedbacks will be studied. It is shown that a beam phase feedback is necessary in order to guarantee a predictable hewn behavior after the first batch injection, otherwise the initial phase deviation for the following batch injections cannot be controlled. However, the effectiveness of the phase feedback control of the transient beam loading may be limited by an emittance blow up in the process. It is shown that a fast power amplifier feedback with a moderate gain can significantly reduce the transient effect of the bunched beam injection

  20. Electron beam solenoid reactor concept

    International Nuclear Information System (INIS)

    The electron Beam Heated Solenoid (EBHS) reactor is a linear magnetically confined fusion device in which the bulk or all of the heating is provided by a relativistic electron beam (REB). The high efficiency and established technology of the REB generator and the ability to vary the coupling length make this heating technique compatible with several radial and axial enery loss reduction options including multiple-mirrors, electrostatic and gas end-plug techniques. This paper addresses several of the fundamental technical issues and provides a current evaluation of the concept. The enhanced confinement of the high energy plasma ions due to nonadiabatic scattering in the multiple mirror geometry indicates the possibility of reactors of the 150 to 300 meter length operating at temperatures > 10 keV. A 275 meter EBHS reactor with a plasma Q of 11.3 requiring 33 MJ of beam eneergy is presented

  1. Critical heat flux (CHF) characteristics of high conversion pressurized water reactor with double flat core

    International Nuclear Information System (INIS)

    Thermal-hydraulic feasibility of a high conversion pressurized water reactor (HCPWR) with a double flat core was studied from a view point of minimum departure from nucleate boiling ratio (MDNBR). The proposed HCPWR improves uranium utilization under the same electrical output as a conventional 3-loop PWR. The critical heat flux (CHF) data of Bettis Atomic Power Laboratory were used to evaluate the predictive capability of KfK, EPRI-B and W and EPRI-Columbia correlations for triangular tight lattice cores. Among them, considering a predictive erroe band of 20%, the KfK correlation had the most promising features for a design purpose. The KfK correlation was applied to COBRA-IV-I steady-state subchannel analysis of the double-flat-core HCPWR under the same boundary conditions as the 3-loop PWR. As a result, it was revealed that the MDNBR of a reference design (fuel rod diameter: 9.5mm, control rod thimble diameter: 11.0mm and rod pitch: 11.4mm) satisfied the lower limit of MDNBR for a current PWR with a sufficient margin. Parametric studies clarified that: (1) The DNBR was increased with increasing the rod pitch or decreasing the control rod thimble diameter. (2) The radial peaking factor was recommended to be less than 1.72 under the present reference core disign. (3) A smaller fuel rod diameter required a larger pitch-to-diameter ratio (p/d) to satisfy the lower limit of MDNBR. Based on the steady-state MDNBR analysis results, the feasibility of the double-flat-core HCPWR was proved under proper considerations of core design and radial power distribution. (author)

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Petrie, Christian M. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); McDuffee, Joel Lee [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Katoh, Yutai [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

    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.

  4. IEEE 1394 CAMERA IMAGING SYSTEM FOR BROOKHAVENS BOOSTER APPLICATION FACILITY BEAM DIAGNOSTICS

    International Nuclear Information System (INIS)

    Brookhaven's Booster Applications Facility (BAF) will deliver resonant extracted heavy ion beams from the AGS Booster to short-exposure fixed-target experiments located at the end of the BAF beam line. The facility is designed to deliver a wide range of heavy ion species over a range of intensities from 103 to over 108 ions/pulse, and over a range of energies from 0.1 to 3.0 GeV/nucleon. With these constraints we have designed instrumentation packages which can deliver the maximum amount of dynamic range at a reasonable cost. Through the use of high quality optics systems and neutral density light filters we will achieve 4 to 5 orders of magnitude in light collection. By using digital IEEE1394 camera systems we are able to eliminate the frame-grabber stage in processing and directly transfer data at maximum rates of 400 Mb/set. In this note we give a detailed description of the system design and discuss the parameters used to develop the system specifications. We will also discuss the IEEE1394 camera software interface and the high-level user interface

  5. Design of a medical reactor generating high quality neutron beams for boron neutron capture therapy

    International Nuclear Information System (INIS)

    Boron neutron capture therapy (BNCT) is a binary treatment modality that can selectively irradiate tumor tissue. BNCT uses drugs containing a stable isotope of boron, B-10, that are capable of preferentially accumulating in the tumor, which is then irradiated with thermal neutrons. The interaction of the B-10 with a thermal neutron causes the B-10 nucleus to split, releasing an alpha particle and a lithium nucleus. These products of the boron neutron capture reaction are very damaging to cells but have a path length in tissue of approximately 14 micrometers, or roughly the diameter of one or two cells. Thus, most of the ionizing energy imparted to tissue is localized to B-10-loaded cells. Since the early 1980s, there have been considerable improvements in boron compounds and neutron beams. More is known now about the radiation biology of BNCT, which has reemerged as a potentially useful method for preferential irradiation of tumors. Clinical trials have been initiated at BNL and MIT, with an improved boron compound and epithermal neutrons. At this time, nuclear reactors are the only demonstrated satisfactory sources of epithermal neutrons. While some reactors are available and within reach of cancer treatment centers, a question arises as to the feasibility and practicality of placing new epithermal neutron sources in hospitals. In this thesis, we design a square reactor (that can easily be reconfigured into polygonal reactors as the need arises) with four slab type assemblies to produce two epithermal neutron beams and two thermal neutron beams for use in neutron capture therapy. This square reactor with four large-area faces consists of 1056 U3Si-Al fuel elements and 36 B4C control rods. The proposed facility, based on this square reactor core with a maximum operating power of 300kW, provides an epithermal neutron beam of 3.2x109 nepi/cm2 · s intensity with low contamination by fast neutrons (<1.6x10-13 Gy · cm2/nepi) and gamma rays (<1.0x10-13 Gy · cm2/nepi

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  7. Neutron Diffraction Patterns Measured with a High-Resolution Powder Diffractometer Installed on a Low-Flux Reactor

    International Nuclear Information System (INIS)

    A powder diffractometer has been recently installed on the IEA-R1 reactor at IPEN-CNEN/SP. IEA-R1 is a light-water open-pool research reactor. At present it operates at 4.5 MW thermal with the possible maximum power of 5 MW. At 4.5 MW the in-core flux is ca. 7x1013 cm-2s-1. In spite of this low flux, installation of both a position-sensitive detector (PSD) and a double-bent silicon monochromator has turned possible to design the new instrument as a high-resolution powder diffractometer. In this work, we present results of the application of the Rietveld method to several neutron powder diffraction patterns. The diffraction patterns were measured in the new instrument with samples of compounds having different structures in order to evaluate the main characteristics of the instrument. (author)

  8. Neutronic flux stability of production uranium graphite reactor conversion core relative to high-frequency oscillations

    International Nuclear Information System (INIS)

    Preliminary methodical simplified investigation into stability of the neutron field in the conversion load of industrial uranium-graphite reactors with regard to basic characteristics of the load in transient processes was carried out. Analysis was based on the calculated research into the behaviour of simplified single-point and one-dimensional models of the reactor core in transient regimes during the interconnected description of dynamics of neutron-physical and thermal properties of the load. Fundamental assumptions on the reactor characteristics used in the calculated model. In the context of accepted approximations the obtained results preclude the possibility for the occurrence of spontaneous high frequency oscillations resulting from the positive reactivity effect on the fuel temperature in the conversion load

  9. Burnout experiments on the externally-finned swirl tube for steady-state and high-heat flux beam stops

    International Nuclear Information System (INIS)

    An experimental study to develop beam stops for the next generation of neutral beam injectors was started, using an ion source developed for the JT-60 neutral beam injector. A swirl tube is one of the most promising candidates for a beam stop element which can handle steady-state and high-heat flux beams. In the present experiments, a modified swirl tube, namely an externally-finned swirl tube, was tested together with a simple smooth tube, an externally finned tube, and an internally finned tube. The major dimensions of the tubes are 10 mm in outer-diameter, 1.5 mm in wall thickness, 15 mm in external fin width, and 700 mm in length. The burnout heat flux (CHF) normal to the externally finned swirl tube was 4.1±0.1 kW/cm2, where the Gaussian e-folding half-width of the beam intensity distribution was about 90 mm, the flow rate of the cooling water was 30 l/min, inlet and outlet gauge pressures were about 1 MPa and 0.2 MPa, respectively, and the temperature of the inlet water was kept to 200C during a pulse. A burnout heat flux ratio, which is defined by the ratio of the CHF value of the externally-finned swirl tube to that of the externally-finned tube, turned out to be about 1.5. Burnout heat fluxes of the tubes with a swirl tape or internal fins increase linearly with an increase of the flow rate. It was found that the tube with external fins has effects that not only reduce the thermal stress but also improve the characteristics of boiling heat transfer. (orig.)

  10. Colliding Beam Fusion Reactors

    Science.gov (United States)

    Rostoker, Norman; Qerushi, Artan; Binderbauer, Michl

    2003-06-01

    The recirculating power for virtually all types of fusion reactors has previously been calculated [1] with the Fokker-Planck equation. The reactors involve non-Maxwellian plasmas. The calculations are generic in that they do not relate to specific confinement devices. In all cases except for a Tokamak with D-T fuel the recirculating power was found to exceed the fusion power by a large factor. In this paper we criticize the generality claimed for this calculation. The ratio of circulating power to fusion power is calculated for the Colliding Beam Reactor with fuels D-T, D-He3 and p-B11. The results are respectively, 0.070, 0.141 and 0.493.

  11. The study on the stability of the supporting ground on the construction site of High Flux Reactor building in Research Reactor Institute of Kyoto University

    International Nuclear Information System (INIS)

    This report provides the results of the study on the stability of the supporting gwound which has been carried out as a part of the seismic design of the High Flux Reactor building which is planned to be constructed by Kyoto University, Research Reactor Institute. In this work the finite element method is used. The stresses and displacements of the ground are calculated under the following conditions; (1) Stress-strain relationships for the individual elements are linear. (2) The problem is analyzed on two-dimensional plane strain distributions. (3) No-tension analysis is applied to the calculation for earthquake load. (4) The mechanical properties of the ground are obtained from the soil survey which has been performed at the construction site of High Flux Reactor building. The results are summarized as follows; (1) The settlement of the building is estimated to be about 2 -- 5 cm for long-time loading, including the result from elastic theory, while the relative settlement is about 0.3 cm at both ends of the building. (2) Safety factor is larger than 1.4 for long-time loading. (3) Maximum angle of the deformation of the building due to the earthquake load is estimated to be about 9.2 x 10-3 degree (1.6 x 10-4 rad). (4) Safety factor is larger than 1.2 -- 1.3 for earthquake load. Judging from these results described above, the ground at the construction site of the High Flux Reactor is appropriate for the supporting ground of the reactor building, and the mat foundation can be adopted for the foundation form. (author)

  12. Analysis of flow reversal of the high flux research reactor frm-ii using the safety code ATHLET

    International Nuclear Information System (INIS)

    In the framework of the reactor safety analysis of the german high flux reactor FRM-II, a design basis accident characterized by loss of shutdown pumps after scram has been investigated, using the thermal hydraulic safety code ATHLET, in order to estimate the mechanical and thermal stresses caused by the spontaneous evaporation of coolant as consequent of flow reversal from forced downward to natural upward circulation. ATHLET is being developed by the german society for reactor safety (GRS) for the application in the safety analysis of power reactors and has been extended and verified and verified to cover the application range for safety analysis of research reactors. The results of simulation show that at the moment of flow stagnation directly before the onset of flow reversal accompanying with a suddenly evaporation of coolant is occurred with significant void increasing to more than 90% causing a suddenly pressure peak of more than 6 bar at the channel outlet beginning with 2 bar, which induces mechanical shock affecting the fuel element structure

  13. High-flux low-divergence positron beam generation from ultra-intense laser irradiated a tapered hollow target

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Jian-Xun [College of Science, National University of Defense Technology, Changsha 410073 (China); College of Electronic Engineering, Wuhan 430019 (China); Ma, Yan-Yun, E-mail: yanyunma@126.com [College of Science, National University of Defense Technology, Changsha 410073 (China); IFSA Collaborative Innovation Center, Shanghai Jiao Tong University, Shanghai 200240 (China); Laser Fusion Research Center, China Academy of Engineering Physics, Mianyang 621000 (China); Zhao, Jun; Yu, Tong-Pu, E-mail: tongpu@nudt.edu.cn; Yang, Xiao-Hu; Gan, Long-Fei; Zhang, Guo-Bo; Yan, Jian-Feng; Zhuo, Hong-Bin; Liu, Jin-Jin; Zhao, Yuan [College of Science, National University of Defense Technology, Changsha 410073 (China); Kawata, Shigeo [Center for Optical Research and Education, Graduate School of Engineering, Utsunomiya University, 7-1-2 Yohtoh, Utsunomiya 321-8585 (Japan)

    2015-10-15

    By using two-dimensional particle-in-cell simulations, we demonstrate high-flux dense positrons generation by irradiating an ultra-intense laser pulse onto a tapered hollow target. By using a laser with an intensity of 4 × 10{sup 23 }W/cm{sup 2}, it is shown that the Breit-Wheeler process dominates the positron production during the laser-target interaction and a positron beam with a total number >10{sup 15} is obtained, which is increased by five orders of magnitude than in the previous work at the same laser intensity. Due to the focusing effect of the transverse electric fields formed in the hollow cone wall, the divergence angle of the positron beam effectively decreases to ∼15° with an effective temperature of ∼674 MeV. When the laser intensity is doubled, both the positron flux (>10{sup 16}) and temperature (963 MeV) increase, while the divergence angle gets smaller (∼13°). The obtained high-flux low-divergence positron beam may have diverse applications in science, medicine, and engineering.

  14. High-flux low-divergence positron beam generation from ultra-intense laser irradiated a tapered hollow target

    International Nuclear Information System (INIS)

    By using two-dimensional particle-in-cell simulations, we demonstrate high-flux dense positrons generation by irradiating an ultra-intense laser pulse onto a tapered hollow target. By using a laser with an intensity of 4 × 1023 W/cm2, it is shown that the Breit-Wheeler process dominates the positron production during the laser-target interaction and a positron beam with a total number >1015 is obtained, which is increased by five orders of magnitude than in the previous work at the same laser intensity. Due to the focusing effect of the transverse electric fields formed in the hollow cone wall, the divergence angle of the positron beam effectively decreases to ∼15° with an effective temperature of ∼674 MeV. When the laser intensity is doubled, both the positron flux (>1016) and temperature (963 MeV) increase, while the divergence angle gets smaller (∼13°). The obtained high-flux low-divergence positron beam may have diverse applications in science, medicine, and engineering

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  16. Response of aluminium and its alloys to exposure in the high flux isotope reactor

    International Nuclear Information System (INIS)

    Pure aluminum and some aluminum alloys were irradiated to very high neutron fluences in the cooling water at 328 K in the high flux region of HFIR. Displacement levels of 270 dpa and transmutation-produced silicon levels of 7.15 wt% were reached. Damage microstructures consisted of dislocations, cavities and precipitates which caused substantial strengthening and associated loss in ductility. Formation of cavities and related swelling were considerably reduced by alloying elements and by the presence of fine Mg2Si precipitate. (author)

  17. Reactor - and accelerator-based filtered beams

    International Nuclear Information System (INIS)

    The neutrons produced in high flux nuclear reactors and in accelerator, induced fission and spallation reactions, represent the most intense sources of neutrons available for research. However, the neutrons from these sources are not monoenergetic, covering the broad range extending from 10-3 eV up to 107 eV or so. In order to make quantitative measurements of the effects of neutrons and their dependence on neutron energy it is desirable to have mono-energetic neutron sources. The paper describes briefly methods of obtaining mono-energetic neutrons and different methods of filtration. This is followed by more detailed discussion of neutron window filters and a summary of the filtered beam facilities using this technique. The review concludes with a discussion of the main applications of filtered beams and their present and future importance

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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

  19. DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2007

    International Nuclear Information System (INIS)

    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

  20. Successful removal of more than 200 spent HEU fuel assemblies from the high flux reactor at Petten

    International Nuclear Information System (INIS)

    Earlier this year, a shipment of 210 spent Highly Enriched Uranium (HEU) Material Test Reactor (MTR) fuel assemblies representing more than 100kg of HEU (pre-irradiation) were shipped from the High Flux Reactor (HFR) in Petten, Netherlands to the US Department of Energy's Savannah River Site. The HFR is owned by the European Commission, Joint Research Centre (JRC) while the Nuclear Research and consultancy Group (NRG) is the reactor operator and license holder. Since the start of the Foreign Research Reactor Program in 1996, this was the largest shipment of MTR fuel elements originating from a single facility. A total of five NAC-LWT casks were used to support this shipment. Multiple technical, licensing, interface and scheduling challenges needed to be overcome to meet the project objectives. During the five months of project planning and execution, the project team (JRC, NRG, Transrad, NAC, DOE and Westinghouse) has been working in close cooperation and good spirit to achieve a safe and successful shipment. The paper presents how the team completed this important shipment for the future of the HFR. (author)

  1. Risk and safety analysis in support of the operation at the High Flux Isotope Reactor at Oak Ridge

    International Nuclear Information System (INIS)

    The High Flux Isotope Reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988 a Level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 x 10-4. It was dominated by flow blockages and loss of all AC power. In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic, wind, and fires. The overall external event contribution to core damage frequency is about 50 percent of the internal event initiated contribution and is dominated by seismic events. Several design and safety analysis studies were undertaken to support the restart of the HFIR. The first study involved a fracture mechanics analysis and redesign of the reactor operating conditions and safety system setting to provide a basis for future operation. Another study involved performing a risk analysis by combining the Level 1 PRA results with offsite consequence analyses under conservative assumptions about the fission product removal within the plant. Additional studies were performed to establish a long-term decay heat removal design basis. Finally, updated and upgraded loss-of-cooling accident studies were performed and are still underway. 5 refs., 11 figs., 6 tabs

  2. Risk and safety analysis in support of the operation at the High Flux Isotope Reactor at oak ridge

    International Nuclear Information System (INIS)

    The High Flux Isotope Reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of the discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988 a Level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 x 10-4. It was dominated by flow blockages and loss of all AC power. In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic events, wind, and fires. The overall external event contribution to core damage frequency is about 50 percent of the internal event initiated contribution and is dominated by seismic events. Several design and safety analysis studies were undertaken to support the restart of the HFIR. The first study involved a fracture mechanics analysis and redesign of the reactor operating conditions and safety system settings to provide a basis for future operation. Another study involved performing a risk analysis by combining the Level 1 PRA results with offsite consequences analysis under conservative assumptions about the fission product removal within the plant. Additional studies were performed to establish a long-term decay heat removal design basis. Finally, updated and upgraded loss-of-cooling accident studies were performed and are still underway

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

    International Nuclear Information System (INIS)

    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/m2 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 has been 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. (J.P.N.) 62 refs

  4. A new deflection technique applied to an existing scheme of electrostatic accelerator for high energy neutral beam injection in fusion reactor devices

    Science.gov (United States)

    Pilan, N.; Antoni, V.; De Lorenzi, A.; Chitarin, G.; Veltri, P.; Sartori, E.

    2016-02-01

    A scheme of a neutral beam injector (NBI), based on electrostatic acceleration and magneto-static deflection of negative ions, is proposed and analyzed in terms of feasibility and performance. The scheme is based on the deflection of a high energy (2 MeV) and high current (some tens of amperes) negative ion beam by a large magnetic deflector placed between the Beam Source (BS) and the neutralizer. This scheme has the potential of solving two key issues, which at present limit the applicability of a NBI to a fusion reactor: the maximum achievable acceleration voltage and the direct exposure of the BS to the flux of neutrons and radiation coming from the fusion reactor. In order to solve these two issues, a magnetic deflector is proposed to screen the BS from direct exposure to radiation and neutrons so that the voltage insulation between the electrostatic accelerator and the grounded vessel can be enhanced by using compressed SF6 instead of vacuum so that the negative ions can be accelerated at energies higher than 1 MeV. By solving the beam transport with different magnetic deflector properties, an optimum scheme has been found which is shown to be effective to guarantee both the steering effect and the beam aiming.

  5. A new deflection technique applied to an existing scheme of electrostatic accelerator for high energy neutral beam injection in fusion reactor devices

    Energy Technology Data Exchange (ETDEWEB)

    Pilan, N., E-mail: nicola.pilan@igi.cnr.it; Antoni, V.; De Lorenzi, A.; Chitarin, G.; Veltri, P.; Sartori, E. [Consorzio RFX—Associazione EURATOM-ENEA per la Fusione, Corso Stati Uniti 4, 35127 Padova (Italy)

    2016-02-15

    A scheme of a neutral beam injector (NBI), based on electrostatic acceleration and magneto-static deflection of negative ions, is proposed and analyzed in terms of feasibility and performance. The scheme is based on the deflection of a high energy (2 MeV) and high current (some tens of amperes) negative ion beam by a large magnetic deflector placed between the Beam Source (BS) and the neutralizer. This scheme has the potential of solving two key issues, which at present limit the applicability of a NBI to a fusion reactor: the maximum achievable acceleration voltage and the direct exposure of the BS to the flux of neutrons and radiation coming from the fusion reactor. In order to solve these two issues, a magnetic deflector is proposed to screen the BS from direct exposure to radiation and neutrons so that the voltage insulation between the electrostatic accelerator and the grounded vessel can be enhanced by using compressed SF{sub 6} instead of vacuum so that the negative ions can be accelerated at energies higher than 1 MeV. By solving the beam transport with different magnetic deflector properties, an optimum scheme has been found which is shown to be effective to guarantee both the steering effect and the beam aiming.

  6. A new deflection technique applied to an existing scheme of electrostatic accelerator for high energy neutral beam injection in fusion reactor devices

    International Nuclear Information System (INIS)

    A scheme of a neutral beam injector (NBI), based on electrostatic acceleration and magneto-static deflection of negative ions, is proposed and analyzed in terms of feasibility and performance. The scheme is based on the deflection of a high energy (2 MeV) and high current (some tens of amperes) negative ion beam by a large magnetic deflector placed between the Beam Source (BS) and the neutralizer. This scheme has the potential of solving two key issues, which at present limit the applicability of a NBI to a fusion reactor: the maximum achievable acceleration voltage and the direct exposure of the BS to the flux of neutrons and radiation coming from the fusion reactor. In order to solve these two issues, a magnetic deflector is proposed to screen the BS from direct exposure to radiation and neutrons so that the voltage insulation between the electrostatic accelerator and the grounded vessel can be enhanced by using compressed SF6 instead of vacuum so that the negative ions can be accelerated at energies higher than 1 MeV. By solving the beam transport with different magnetic deflector properties, an optimum scheme has been found which is shown to be effective to guarantee both the steering effect and the beam aiming

  7. High-Speed Neutron and Gamma Flux Sensor for Monitoring Surface Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA needs compact nuclear reactors to power future bases on the moon and Mars. These reactors require robust automatic control systems using low mass, rapid...

  8. High-Speed Neutron and Gamma Flux Sensor for Monitoring Surface Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA needs compact nuclear reactors to power future bases on the moon and/or Mars. These reactors require robust automatic control systems using low mass, rapid...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  10. Numerical Simulation of Subcooled Boiling Inside High-Heat-Flux Component with Swirl Tube in Neutral Beam Injection System

    International Nuclear Information System (INIS)

    In order to realize steady-state operation of the neutral beam injection (NBI) system with high beam energy, an accurate thermal analysis and a prediction about working conditions of heat-removal structures inside high-heat-flux (HHF) components in the system are key issues. In this paper, taking the HHF ion dump with swirl tubes in NBI system as an example, an accurate thermal dynamic simulation method based on computational fluid dynamics (CFD) and the finite volume method is presented to predict performance of the HHF component. In this simulation method, the Eulerian multiphase method together with some empirical corrections about the inter-phase transfer model and the wall heat flux partitioning model are considered to describe the subcooled boiling. The reliability of the proposed method is validated by an experimental example with subcooled boiling inside swirl tube. The proposed method provides an important tool for the refined thermal and flow dynamic analysis of HHF components, and can be extended to study the thermal design of other complex HHF engineering structures in a straightforward way. The simulation results also verify that the swirl tube is a promising heat removing structure for the HHF components of the NBI system. (fusion engineering)

  11. Brookhaven highlights. [Fiscal year 1992, October 1, 1991--September 30, 1992

    Energy Technology Data Exchange (ETDEWEB)

    Rowe, M.S.; Cohen, A.; Greenberg, D.; Seubert, L. [eds.

    1992-12-31

    This publication provides a broad overview of the research programs and efforts being conducted, built, designed, and planned at Brookhaven National Laboratory. This work covers a broad range of scientific disciplines. Major facilities include the Alternating Gradient Synchrotron (AGS), with its newly completed booster, the National Synchrotron Light Source (NSLS), the High Flux Beam Reactor (HFBR), and the RHIC, which is under construction. Departments within the laboratory include the AGS department, accelerator development, physics, chemistry, biology, NSLS, medical, nuclear energy, and interdepartmental research efforts. Research ranges from the pure sciences, in nuclear physics and high energy physics as one example, to environmental work in applied science to study climatic effects, from efforts in biology which are a component of the human genome project to the study, production, and characterization of new materials. The paper provides an overview of the laboratory operations during 1992, including staffing, research, honors, funding, and general laboratory plans for the future.

  12. Development of a high temperature, high sensitivity fission counter for liquid metal reactor in-vessel flux monitoring

    International Nuclear Information System (INIS)

    Advanced liquid metal reactor concepts such as the Sodium Advanced Fast Reactor (SAFR) and the Power Reactor Inherently Safe Module (PRISM) have relatively large pressure vessels that necessitate in-vessel placement of the neutron detectors to achieve adequate count rates during source range operations. It is estimated that detector sensitivities of 5 to 10 counts/center dot/s/center dot//sup /minus/1//center dot/[neutron/(cm2/center dot/s)]/sup /minus/1/ will be required for the initial core loading. The Instrumentation and Controls Division of Oak Ridge National Laboratory has designed and fabricated a fission counter to meet this requirement which is also capable of operating in uncooled instrument thimbles at primary coolant temperatures of 500 to 600/degree/C. Components are fabricated from Inconel-600, and high temperature alumina insulators are employed. The transmission line electrode configuration is utilized to minimize capacitive loading effects

  13. BR2 reactor neutron beams

    International Nuclear Information System (INIS)

    The use of reactor neutron beams is becoming increasingly more widespread for the study of some properties of condensed matter. It is mainly due to the unique properties of the ''thermal'' neutrons as regards wavelength, energy, magnetic moment and overall favorable ratio of scattering to absorption cross-sections. Besides these fundamental reasons, the impetus for using neutrons is also due to the existence of powerful research reactors (such as BR2) built mainly for nuclear engineering programs, but where a number of intense neutron beams are available at marginal cost. A brief introduction to the production of suitable neutron beams from a reactor is given. (author)

  14. Chronology of the beryllium replacement shutdown at the High Flux Isotope Reactor (HFIR), 1983

    International Nuclear Information System (INIS)

    In addition to the permanent beryllium reflector, several other components were replaced. The outer shroud and lower tracks were replaced. The new control rod access plugs and the upper tracks were installed. Replacement of collimator tubes for HB-1 and -2 are tentatively slated for the next permanent beryllium changeout. Inspection of the reactor vessel, the vessel-to-nozzle welds, core support structure, and vessel internal cladding showed them to be in acceptable condition. The highest, accumulative radiation doses received by Reactor Operations personnel during the shutdown, in mrem, were 665, 606, and 560; the highest for P and E personnel were 520, 505, and 475

  15. Reliable operation of the Brookhaven EBIS for highly charged ion production for RHIC and NSRL

    Science.gov (United States)

    Beebe, E.; Alessi, J.; Binello, S.; Kanesue, T.; McCafferty, D.; Morris, J.; Okamura, M.; Pikin, A.; Ritter, J.; Schoepfer, R.

    2015-01-01

    An Electron Beam Ion Source for the Relativistic Heavy Ion Collider (RHIC EBIS) was commissioned at Brookhaven in September 2010 and since then it routinely supplies ions for RHIC and NASA Space Radiation Laboratory (NSRL) as the main source of highly charged ions from Helium to Uranium. Using three external primary ion sources for 1+ injection into the EBIS and an electrostatic injection beam line, ion species at the EBIS exit can be switched in 0.2 s. A total of 16 different ion species have been produced to date. The length and the capacity of the ion trap have been increased by 20% by extending the trap by two more drift tubes, compared with the original design. The fraction of Au32+ in the EBIS Au spectrum is approximately 12% for 70-80% electron beam neutralization and 8 pulses operation in a 5 Hertz train and 4-5 s super cycle. For single pulse per super cycle operation and 25% electron beam neutralization, the EBIS achieves the theoretical Au32+ fractional output of 18%. Long term stability has been very good with availability of the beam from RHIC EBIS during 2012 and 2014 RHIC runs approximately 99.8%.

  16. Reliable operation of the Brookhaven EBIS for highly charged ion production for RHIC and NSRL

    International Nuclear Information System (INIS)

    An Electron Beam Ion Source for the Relativistic Heavy Ion Collider (RHIC EBIS) was commissioned at Brookhaven in September 2010 and since then it routinely supplies ions for RHIC and NASA Space Radiation Laboratory (NSRL) as the main source of highly charged ions from Helium to Uranium. Using three external primary ion sources for 1+ injection into the EBIS and an electrostatic injection beam line, ion species at the EBIS exit can be switched in 0.2 s. A total of 16 different ion species have been produced to date. The length and the capacity of the ion trap have been increased by 20% by extending the trap by two more drift tubes, compared with the original design. The fraction of Au32+ in the EBIS Au spectrum is approximately 12% for 70-80% electron beam neutralization and 8 pulses operation in a 5 Hertz train and 4-5 s super cycle. For single pulse per super cycle operation and 25% electron beam neutralization, the EBIS achieves the theoretical Au32+ fractional output of 18%. Long term stability has been very good with availability of the beam from RHIC EBIS during 2012 and 2014 RHIC runs approximately 99.8%

  17. Beam characterization at the Neutron Radiography Reactor

    International Nuclear Information System (INIS)

    Highlights: • The project characterized the beam at the Neutron Radiography Reactor. • Experiments indicate that the neutron energy spectrum model may not be accurate. • The facility is a category I radiography facility. • The beam divergence and effective collimation ratio are 0.3 ± 0.1° and >125. • The predicted total neutron flux at the image plane is 5.54 × 106 n/cm2 s. -- Abstract: The quality of a neutron-imaging beam directly impacts the quality of radiographic images produced using that beam. Fully characterizing a neutron beam, including determination of the beam's effective length-to-diameter ratio, neutron flux profile, energy spectrum, potential image quality, and beam divergence, is vital for producing quality radiographic images. This paper provides a characterization of the east neutron imaging beamline at the Idaho National Laboratory Neutron Radiography Reactor (NRAD). The experiments which measured the beam's effective length-to-diameter ratio and potential image quality are based on American Society for Testing and Materials (ASTM) standards. An analysis of the image produced by a calibrated phantom measured the beam divergence. The energy spectrum measurements consist of a series of foil irradiations using a selection of activation foils, compared to the results produced by a Monte Carlo n-Particle (MCNP) model of the beamline. The NRAD has an effective collimation ratio greater than 125, a beam divergence of 0.3 ± 0.1°, and a gold foil cadmium ratio of 2.7. The flux profile has been quantified and the facility is an ASTM Category 1 radiographic facility. Based on bare and cadmium covered foil activation results, the neutron energy spectrum used in the current MCNP model of the radiography beamline over-samples the thermal region of the neutron energy spectrum

  18. TIBER II [Tokamak Ignition/Burn Experimental Reactor] parameters with neutral beams at high energies

    International Nuclear Information System (INIS)

    The baseline neutral beam energy for TIBER II was chosen to be 500 keV consistent with the use of near term dc acceleration technology. Adequate penetration to the axis for core current drive in larger ETR devices requires higher beam energies. However, beam instabilities may limit the current drive efficiency at high energy to lower values than predicted classically. The characteristics of TIBER II and a device with 4.5 m major radius as functions of beam energy are presented. 11 refs

  19. High flux materials testing reactor HFR Petten. Characteristics of facilities and standard irradiation devices

    International Nuclear Information System (INIS)

    For the materials testing reactor HFR some characteristic information is presented. Besides the nuclear data for the experiment positions short descriptions are given of the most important standard facilities for material irradiation and radionuclide production. One paragraph deals with the experimental set-ups for solid state and nuclear structure investigations. The information in this report refers to a core type, which is operational since March 1977. The numerical data compiled have been up-dated to June 1978

  20. Development of control rod driving mechanism for high neutron flux reactor in Kyoto University (KUHFR)

    International Nuclear Information System (INIS)

    KUHFR is a coupling type reactor of 30 MW power output, which have two light-water-moderated and cooled cores inside the heavy water reflector. There are six sets of control rod driving mechanism (CRDM) in each core, each set driving one control rod. The newly developed driving system for CRDM is a unique one not employed in any other reactor. The main specifications required are as follows: Drive length 650 mm, driving speed 100 mm/min; control rod magnet deenergizing time 0.3 sec or less, control rod falling time to 90% stroke 1 sec or less, finished O.D. 190 mm or less. There were difficulties in selecting the driving system, because various control rod driving systems adopted in power and research reactors have both merits and demerits. As a result of investigation, three systems have been produced for trial, experimented and compared, and the moving coil type CRDM has been employed because it is suitable in many points, e.g. it allows continuous motion of control rods. The construction of moving coil type CRDM is explained. In the progress of development from No. 1 to No. 3 system is described, starting at the magnetic circuit calculation. As the running performance of the CRDM, the relationship between the plunger shift in a coil and upward force, and the differential linear running performance, following properties and stopping characteristics of control rods for coil movement are described. (Wakatsuki, Y.)

  1. 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)

    Energy Technology Data Exchange (ETDEWEB)

    Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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.

  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)

    International Nuclear Information System (INIS)

    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. Beam emittance and output waveforms of high-flux laser ion source

    International Nuclear Information System (INIS)

    A laser ion source with short drift distance has been developed for a driver of heavy ion fusion (HIF). It supplies a copper ion beam of 200 mA (255 mA/cm2) with duration of 400 ns and beam emittance is about 0.8π mm·mrad. Moreover it has fast rising (30 ns), flat-top current waveform and a potential to deliver pure charge states between 1+ - 3+. Experimental results indicate that the laser ion source is a good candidate for the HIF driver. (author)

  4. Brookhaven highlights

    International Nuclear Information System (INIS)

    This report highlights the research activities of Brookhaven National Laboratory during the period dating from October 1, 1992 through September 30, 1993. There are contributions to the report from different programs and departments within the laboratory. These include technology transfer, RHIC, Alternating Gradient Synchrotron, physics, biology, national synchrotron light source, applied science, medical science, advanced technology, chemistry, reactor physics, safety and environmental protection, instrumentation, and computing and communications

  5. Lessons Learned from the Application of Bulk Characterization to Individual Containers on the Brookhaven Graphite Research Reactor Decommissioning Project at Brookhaven National Laboratory - 12056

    International Nuclear Information System (INIS)

    When conducting environmental cleanup or decommissioning projects, characterization of the material to be removed is often performed when the material is in-situ. The actual demolition or excavation and removal of the material can result in individual containers that vary significantly from the original bulk characterization profile. This variance, if not detected, can result in individual containers exceeding Department of Transportation regulations or waste disposal site acceptance criteria. Bulk waste characterization processes were performed to initially characterize the Brookhaven Graphite Research Reactor (BGRR) graphite pile and this information was utilized to characterize all of the containers of graphite. When the last waste container was generated containing graphite dust from the bottom of the pile, but no solid graphite blocks, the material contents were significantly different in composition from the bulk waste characterization. This error resulted in exceedance of the disposal site waste acceptance criteria. Brookhaven Science Associates initiated an in-depth investigation to identify the root causes of this failure and to develop appropriate corrective actions. The lessons learned at BNL have applicability to other cleanup and demolition projects which characterize their wastes in bulk or in-situ and then extend that characterization to individual containers. (authors)

  6. Lessons Learned from the Application of Bulk Characterization to Individual Containers on the Brookhaven Graphite Research Reactor Decommissioning Project at Brookhaven National Laboratory - 12056

    Energy Technology Data Exchange (ETDEWEB)

    Kneitel, Terri [US DOE, Brookhaven Site Office (United States); Rocco, Diane [Brookhaven National Laboratory (United States)

    2012-07-01

    When conducting environmental cleanup or decommissioning projects, characterization of the material to be removed is often performed when the material is in-situ. The actual demolition or excavation and removal of the material can result in individual containers that vary significantly from the original bulk characterization profile. This variance, if not detected, can result in individual containers exceeding Department of Transportation regulations or waste disposal site acceptance criteria. Bulk waste characterization processes were performed to initially characterize the Brookhaven Graphite Research Reactor (BGRR) graphite pile and this information was utilized to characterize all of the containers of graphite. When the last waste container was generated containing graphite dust from the bottom of the pile, but no solid graphite blocks, the material contents were significantly different in composition from the bulk waste characterization. This error resulted in exceedance of the disposal site waste acceptance criteria. Brookhaven Science Associates initiated an in-depth investigation to identify the root causes of this failure and to develop appropriate corrective actions. The lessons learned at BNL have applicability to other cleanup and demolition projects which characterize their wastes in bulk or in-situ and then extend that characterization to individual containers. (authors)

  7. Measurement of (n,xn) reaction cross-sections using prompt γ spectroscopy at neutron beams with high instantaneous flux

    International Nuclear Information System (INIS)

    The work presented in this thesis is situated in the context of the GEDEON program of neutron cross-section measurements. This program is motivated by the perspectives recently opened by projects of nuclear waste treatment and energy production. There is an obvious lack of experimental data on (n,xn) reactions in the databases, especially in the case of very radioactive isotopes. An important technique to measure cross-sections of these reactions is the prompt γ-ray spectroscopy at white pulsed neutron beams with very high instantaneous flux. In this work, inelastic scattering and (n,xn) reactions cross-section measurements were performed on a lead sample from threshold to 20 MeV by prompt γ-ray spectroscopy at the white neutron beam generated by GELINA facility in Geel, Belgium. Digital methods were developed to treat HPGe CLOVER detector signals and separate γ-rays induced by the fastest neutrons from those belonging to the flash. Partial cross-sections for the production of several transitions in natural lead were measured and analyzed using theoretical calculations in order to separate the contributions of different reactions leading to the same residual isotope. Total cross-sections of the reactions in question were estimated. The results were compared to the TALYSS code theoretical calculations, as well as to other experimental results. This experiment has served to validate the method and it opens the way to measure (n,xn) reactions cross-sections with high instantaneous neutron flux on actinides, particularly the U233(n,2n) reaction which is important for the thorium cycle. (author)

  8. Neutron spectra at two beam ports of a TRIGA Mark III reactor loaded with HEU fuel

    International Nuclear Information System (INIS)

    The neutron spectra have been measured in two beam ports, one radial and another tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research in Mexico. Measurements were carried out with the reactor core loaded with high enriched uranium fuel. Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a 6LiI(Eu) scintillator and 2, 3, 5, 8, 10 and 12 in.-diameter high-density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code. For each spectrum total flux, mean energy and ambient dose equivalent were determined. Measured spectra show fission, epithermal and thermal neutrons, being harder in the radial beam port. - Highlights: • Neutron spectra of a TRIGA reactor were measured. • The reactor core is loaded with HEU. • The spectra were measured at two reactor beam ports. • Measurements were carried out at 5 and 10 W

  9. Pursuing nuclear energy with no nuclear contamination - from neutron flux reactor to deuteron flux reactor

    International Nuclear Information System (INIS)

    -free channel in the deuteron-deuteron fusion reaction. Even if polarized deuteron has long enough life-time to keep its polarity in hot fusion plasma, there is still the probability to have the neutron emission channel from deuteron-deuteron fusion. The neutron emission in hot plasma containing deuterons is inevitable. Isomer Hf-178 was proposed to reduce the neutron emission in terms of gamma decay controlled by X-ray. Although its reality is still in question, the resonance plays key role in this concept as well. Condensed matter nuclear science provided another chance to approach nuclear energy with no nuclear contamination. Selective resonant tunneling would select only the neutron free channel. There are five major steps in the past 17 years: (1) Selective Resonant tunneling model has been successful to explain the 3 major puzzles in cold fusion proposed by nuclear physicist(i.e. penetration of Coulomb barrier, no neutron emission, no gamma radiation), and successful also to explain the 3 major cross-section data in hot fusion(i.e. d+t, d+d, d+He3). The Nobel prize laureate, B. Josephson of Cambridge University, cited this theory in the famous Lindau Meeting (2004).[1,2] (2) Deuteron flux through the palladium surface at specific temperature was found correlated with heat flow in various experiments in China, Switzerland, Japan, France and Italy.[3,4] (3) The nuclear products have been confirmed in a series of nuclear transmutation experiments using deuterium flux permeating through the thin film on the palladium surface.[5] (4) Distinct from the beam-target experiment, a special procedure was proposed to search this resonance between lattice energy level and nuclear energy level. (5) Instead of the electrolytic cell, the gas loading technique has been used. It led to the discovery of the temperature of these resonances which may be as high as 1000 degree C. This would change greatly the usage of this nuclear energy. We may propose the future subjects of study as follows

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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 U3O8 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

  12. Validation of a Monte Carlo based depletion methodology via High Flux Isotope Reactor HEU post-irradiation examination measurements

    International Nuclear Information System (INIS)

    The purpose of this study is to validate a Monte Carlo based depletion methodology by comparing calculated post-irradiation uranium isotopic compositions in the fuel elements of the High Flux Isotope Reactor (HFIR) core to values measured using uranium mass-spectrographic analysis. Three fuel plates were analyzed: two from the outer fuel element (OFE) and one from the inner fuel element (IFE). Fuel plates O-111-8, O-350-I, and I-417-24 from outer fuel elements 5-O and 21-O and inner fuel element 49-I, respectively, were selected for examination. Fuel elements 5-O, 21-O, and 49-I were loaded into HFIR during cycles 4, 16, and 35, respectively (mid to late 1960s). Approximately one year after each of these elements were irradiated, they were transferred to the High Radiation Level Examination Laboratory (HRLEL) where samples from these fuel plates were sectioned and examined via uranium mass-spectrographic analysis. The isotopic composition of each of the samples was used to determine the atomic percent of the uranium isotopes. A Monte Carlo based depletion computer program, ALEPH, which couples the MCNP and ORIGEN codes, was utilized to calculate the nuclide inventory at the end-of-cycle (EOC). A current ALEPH/MCNP input for HFIR fuel cycle 400 was modified to replicate cycles 4, 16, and 35. The control element withdrawal curves and flux trap loadings were revised, as well as the radial zone boundaries and nuclide concentrations in the MCNP model. The calculated EOC uranium isotopic compositions for the analyzed plates were found to be in good agreement with measurements, which reveals that ALEPH/MCNP can accurately calculate burn-up dependent uranium isotopic concentrations for the HFIR core. The spatial power distribution in HFIR changes significantly as irradiation time increases due to control element movement. Accurate calculation of the end-of-life uranium isotopic inventory is a good indicator that the power distribution variation as a function of space and

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  14. Thermal neutron flux distribution in ET-RR-2 reactor thermal column

    OpenAIRE

    Imam Mahmoud M.; Roushdy Hassan

    2002-01-01

    The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a) to provide a thermal neutron flux in the neutron transmutation silicon doping, (b) to provide a thermal flux in the neutron activation analysis position, and (c) to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, ...

  15. Status of high current R and D Energy Recovery LINAC at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    An ampere class 20 MeV superconducting Energy Recovery Linac (ERL) is under construction at Brookhaven National Laboratory (BNL) for testing of concepts relevant for high-energy coherent electron cooling and electron-ion colliders. One of the goals is to demonstrate an electron beam with high charge per bunch (∼ 5 nC) and low normalized emittance (∼ 5 mm-mrad) at an energy of 20 MeV. Flexible lattice of ERL loop provides a test-bed for investigating issues of transverse and longitudinal instabilities, and diagnostics for intense CW e-beam. The superconducting 703 MHz RF photoinjector is considered as an electron source for such a facility. We will start with a straight pass (gun - 5 cell cavity - beam stop) test for the SRF Gun performance studies. Later, we will install and test a novel injection line concept for emittance preservation in a lower energy merger. In this paper we present the status and our plans for construction and commissioning of this facility.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Calzavara, Y. [Inst. Laue-Langevin (ILL), Grenoble (France)

    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.

  17. Status of the conversion working plan in the High Flux Reactor (Petten, The Netherlands)

    International Nuclear Information System (INIS)

    The conversion from HEU to LEU has often many disadvantages: flux penalties, increase of fuel consumption, cost and delay to obtain a new license etc. But to fulfill the non-proliferation programme, and to simplify the future fuel supply, the HFR renewed in 1998 studies on conversion possibilities. To minimize the conversion costs, these studies were made with a progressive conversion that avoids the need of one new core and permits to begin the conversion with a replacement of 5 elements at each cycle. Hence the conversion can be made in 7 cycles, without special elements and with a normal bum-up for each element. To avoid an increase of fuel consumption, an increase of the fuel cycle length from 24.7 to 28.3 days was also considered. This point allows reducing the number of annual cycles from 1 to 10 and enables in one cycle to have the possibility of four successive irradiations for Molybdenum production (7 days) in one irradiation position. A working plan for fuel licensing has been sent to the safety authorities and is presented in the paper. (author)

  18. Thermal neutron flux distribution in ET-RR-2 reactor thermal column

    Directory of Open Access Journals (Sweden)

    Imam Mahmoud M.

    2002-01-01

    Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

  19. Radiation Dosimetry in the BNCT Patient Treatment Room at the Brookhaven Medical Research Reactor

    International Nuclear Information System (INIS)

    The BMRR was a 3 MW light water reactor that had an epithermal neutron beam that was used to perform clinical trials on patients with malignant brain tumors. A series of measurements and calculations had been performed in the treatment room both prior to the trials and during the trials. The details of the measurements and the Monte Carlo calculations are presented and compared.

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

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, David [ORNL

    2014-03-01

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

  1. Measurements of neutron flux in the RA reactor

    International Nuclear Information System (INIS)

    This report includes results of the following measurements performed at the RA reactor: thermal neutron flux in the experimental channels, epithermal and fast neutron flux, neutron flux in the biological shield, neutron flux distribution in the reactor cell

  2. 177Lu radiochemical separation from 176Yb irradiated in high-flux research reactor SM

    International Nuclear Information System (INIS)

    Ytterbium and lutetium behaviour has been studied during electrolysis of aqueous solutions containing their chlorides and alkali metal citrate (Li, Na, K). The conditions providing the efficient extraction of ytterbium macro amounts into a mercury-pool cathode have been determined. Laboratory-scale experiments were performed to elaborate chromatographic procedures for 177Lu purification from ytterbium macro amounts and accompanying impurities including hafnium (177Lu radioactive decay product). The conditions providing the efficient separation of 177Lu from the above-mentioned impurities using cation-exchange (in α-hydroxy isobutyric acid) and extraction-chromatographic (impregnated with di-2-ethylhexyl phosphoric acid teflon powder as stationary phase and nitric or hydrochloric acids as eluant) methods have been found. Isotopically enriched ytterbium preparation (176Yb - 95.15; 174Yb - 2.47 atomic %) was purified from lutetium impurity and samples of the purified starting material were irradiated in the central neutron trap and beryllium reflector channel of the SM reactor. 177Lu was extracted from the irradiated targets by electroreduction of ytterbium on the mercury-pool cathode from lithium citrate solution. Cation exchange and extraction chromatography methods were used for subsequent purification of 177Lu. The radiochemical processing took about 50 hours. The results of analysis obtained by the spectrometry of X-ray and gamma radiation, mass-spectrometry and emission spectroscopy are as follows: Chemical form: 177LuCl3, solution in hydrochloric acid; solvent (HCl) concentration: 0.01 - 0.1 mol/l; 177Lu specific activity: ≥ 20 Ci/mg; 177mLu to 177Lu activity ratio: ≤ 0.02 %; total gamma emitters (Co-58, Co-60, Zn-65, Mn-54, Fe-59, Cr-51) to 177Lu activity ratio: ≤ 0.01 %; total mass of non-radioactive impurities (Cu, Zn, Al, Fe, Pb) to 177Lu activity ratio: ≤ 500 μg/Ci; total alpha emitters to 177Lu activity ratio ≤ 1x10-5 %. (author)

  3. SIPHORE: Conceptual Study of a High Efficiency Neutral Beam Injector Based on Photo-detachment for Future Fusion Reactors

    International Nuclear Information System (INIS)

    An innovative high efficiency neutral beam injector concept for future fusion reactors is under investigation (simulation and R and D) between several laboratories in France, the goal being to perform a feasibility study for the neutralization of intense high energy (1 MeV) negative ion (NI) beams by photo-detachment.The objective of the proposed project is to put together the expertise of three leading groups in negative ion quantum physics, high power stabilized lasers and neutral beam injectors to perform studies of a new injector concept called SIPHORE (SIngle gap PHOto-neutralizer energy REcovery injector), based on the photo-detachment of negative ions and energy recovery of unneutralised ions; the main feature of SIPHORE being the relevance for the future Fusion reactors (DEMO), where high injector efficiency (up to 70-80%), technological simplicity and cost reduction are key issues to be addressed.The paper presents the on-going developments and simulations around this project, such as, a new concept of ion source which would fit with this injector topology and which could solve the remaining uniformity issue of the large size ion source, and, finally, the presentation of the R and D program in the laboratories (LAC, ARTEMIS) around the photo-neutralization for Siphore.

  4. Transmutation of /sup 90/Sr and /sup 137/Cs in a high-flux fast reactor with a thermalized central region

    Energy Technology Data Exchange (ETDEWEB)

    Taube, M.

    1976-10-01

    The fission products /sup 90/Sr and /sup 137/Cs produced by fission reactors of 30 GW(th) can be transmutated into stable nuclides by neutron irradiation with a thermal flux of 2 x 10/sup 16/ n cm/sup -2/ s/sup -1/. The rates of transmutation are 15 and 3.3 times greater, respectively, than that of spontaneous beta decay. The transmutation would take place in a central thermalized region of a high-flux fast burner reactor of 7 GW(th). In the case where the power reactors of 23 GW(th) are breeders with a high breeding gain of G = 0.38, the total system, inclusive of the high-flux burner, remains a breeding system, with G/sub total/ = 0.09. Details of the neutronics calculations and simplified thermohydraulics are given. The high-flux burner is fueled with a molten salt of chlorides of plutonium and sodium with a power density of 10 kW cm/sup -3/. The ''self-liquidation'' of such a system is discussed.

  5. Deployment of Smart 3D Subsurface Contaminant Characterization at the Brookhaven Graphite Research Reactor

    International Nuclear Information System (INIS)

    The Brookhaven Graphite Research Reactor (BGRR) Historical Site Assessment (BNL 1999) identified contamination inside the Below Grade Ducts (BGD) resulting from the deposition of fission and activation products from the pile on the inner carbon steel liner during reactor operations. Due to partial flooding of the BGD since shutdown, some of this contamination may have leaked out of the ducts into the surrounding soils. The baseline remediation plan for cleanup of contaminated soils beneath the BGD involves complete removal of the ducts, followed by surveying the underlying and surrounding soils, then removing soil that has been contaminated above cleanup goals. Alternatively, if soil contamination around and beneath the BGD is either non-existent/minimal (below cleanup goals) or is very localized and can be ''surgically removed'' at a reasonable cost, the BGD can be decontaminated and left in place. The focus of this Department of Energy Accelerated Site Technology Deployment (DOE ASTD) project was to determine the extent (location, type, and level) of soil contamination surrounding the BGD and to present this data to the stakeholders as part of the Engineering Evaluation/Cost Analysis (EE/CA) process. A suite of innovative characterization tools was used to complete the characterization of the soil surrounding the BGD in a cost-effective and timely fashion and in a manner acceptable to the stakeholders. The tools consisted of a tracer gas leak detection system that was used to define the gaseous leak paths out of the BGD and guide soil characterization studies, a small-footprint Geoprobe to reach areas surrounding the BGD that were difficult to access, two novel, field-deployed, radiological analysis systems (ISOCS and BetaScint) and a three-dimensional (3D) visualization system to facilitate data analysis/interpretation. All of the technologies performed as well or better than expected and the characterization could not have been completed in the same time or at

  6. Large break loss of coolant severe accident sequences at the HFIR [High Flux Isotope Reactor

    International Nuclear Information System (INIS)

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

  7. Radiological protection considerations during the treatment of glioblastoma patients by boron neutron capture therapy at the high flux reactor in Petten, The Netherlands

    International Nuclear Information System (INIS)

    A clinical trial of Boron Neutron Capture Therapy (BNCT) for glioblastoma patients has been in progress at the High Flux Reactor (HFR) at Petten since October 1997. The JRC (as licence holder of the HFR) must ensure that radiological protection measures are provided. The BNCT trial is a truly European trial, whereby the treatment takes place at a facility in the Netherlands under the responsibility of clinicians from Germany and patients are treated from several European countries. Consequently, radiological protection measures satisfy both German and Dutch laws. To respect both laws, a BNCT radioprotection committee was formed under the chairmanship of an independent radioprotection expert, with members representing all disciplines in the trial. A special nuance of BNCT is that the radiation is provided by a mixed neutron/gamma beam. The radiation dose to the patient is thus a complex mix due to neutrons, gammas and neutron capture in boron, nitrogen and hydrogen, which, amongst others, need to be correctly calculated in non-commercial and validated treatment planning codes. Furthermore, due to neutron activation, measurements on the patient are taken regularly after treatment. Further investigations along these lines include dose determination using TLDs and boron distribution measurements using on-line gamma ray spectroscopy. (author)

  8. EL-2 reactor: Thermal neutron flux distribution

    International Nuclear Information System (INIS)

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  9. The Brookhaven Accelerator Test Facility

    International Nuclear Information System (INIS)

    The Accelerator Test Facility (ATF), presently under construction at Brookhaven National laboratory, is described. It consists of a 50-MeV electron beam synchronizable to a high-peak power CO2 laser. The interaction of electrons with the laser field will be probed, with some emphasis on exploring laser-based acceleration techniques. 5 refs., 2 figs

  10. High field magnet program at Brookhaven National Laboratory

    CERN Document Server

    Ghosh, A; Muratore, J; Parker, B; Sampson, W; Wanderer, P J; Willen, E

    2000-01-01

    The magnet program at Brookhaven National Laboratory (BNL) is focussed on superconducting magnets for particle accelerators. The effort includes magnet production at the laboratory and in industry, magnet R&D, and test facilities for magnets and superconductors. Nearly 2000 magnets-dipoles, quadrupoles, sextupoles and correctors for the arc and insertion regions-were produced for the Relativistic Heavy Ion Collider (RHIC), which is being commissioned. Currently, production of helical dipoles for the polarized proton program at RHIC, insertion region dipoles for the Large Hadron Collider (LHC) at CERN, and an insertion magnet system for the Hadron-Elektron-Ring- Analage (HERA) collider at Deutsches Elektronen-Synchrotron (DESY) is underway. The R&D effort is exploring dipoles with fields above 10 T for use in post-LHC colliders. Brittle superconductors-Nb/sub 3/Sn or HTS-are being used for these magnets. The superconductor test facility measures short-sample currents and other characteristics of sample...

  11. Intermediate energy neutron beams from the MURR [University of Missouri Research Reactor

    International Nuclear Information System (INIS)

    Several reactors in the US are potential candidates to deliver beams of intermediate energy neutrons for NCT. At this time, moderators, as compared to filters, appear to be the more effective means of tailoring the flux of these reactors. The objective is to sufficiently reduce the flux of fast neutrons while producing enough intermediate energy neutrons for treatments. At the University of Missouri Research Reactor (MURR), the code MCNP has recently been used to calculate doses in a phantom. First, ideal beams of 1, 35, and 1,000 eV neutrons were analyzed to determine doses and advantage depths in the phantom. Second, a high quality beam that had been designed to fit in the thermal column of the MURR, was reanalyzed. MCNP calculations of the dose in phantom in this beam confirmed previous calculations and showed that this beam would be a nearly ideal one with neutrons of the desired energy and also a high neutron current. However, installation of this beam will require a significant modification of the thermal column of the MURR. Therefore, a second beam that is less difficult to build and install, but of lower neutron current, has been designed to fit in MURR port F. This beam is designed using inexpensive Al, S, and Pb. The doses calculated in the phantom placed in this beam show that it will be satisfactory for sample tests, animal tests, and possible initial patient trials. Producing this beam will require only modest modifications of the existing tube

  12. Use of neutron beams for low and medium flux research reactors: Radiography and materials characterization. Report of a technical committee held in Vienna, 4-7 May 1993

    International Nuclear Information System (INIS)

    The present report is the result of the Technical Committee meeting held during 4-7 May 1993 in Vienna, Austria, and includes contributions from the participants. The Physics Section of the Department of Research and Isotopes was responsible for the co-ordination and compilation of the report. The report is intended to provide guidelines to research reactor owners and operators for promoting and developing their research programmes and industrial applications for neutron radiology, related neutron inspection and analytical techniques and neutron beam irradiation. Refs, figs and tabs

  13. High Intensity Secondary Beams Driven by Protons

    CERN Document Server

    Galambos, John; Nagaitsev, Sergei

    2013-01-01

    As part of the Intensity Frontier effort within the 2013 Community Summer Study, a workshop on the proton machine capabilities was held (High Intensity Secondary Beams Driven by Proton Beams) April 17-20, 2013 at Brookhaven National Laboratory in Upton, NY. Primary aims of the workshop were to understand: 1) the beam requirements for proposed high intensity proton beam based measurements; 2) the capabilities of existing world-wide high power proton machines; 3) proton facility upgrade plans and proposals for new facilities; 4) and to document the R&D needs for proton accelerators and target systems needed to support proposed intensity frontier measurements. These questions are addressed in this summary.

  14. Design, fabrication, and testing of gadolinium-shielded metal fuel samples in the hydraulic tube of the high flux isotope reactor

    International Nuclear Information System (INIS)

    The use of hydraulic rabbit capsules inserted into and ejected from the core of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) during full power operation allows for precise control of the neutron fluence in fueled experiments. Rabbit capsules with strong thermal neutron absorbers must be used to screen out thermal neutrons, thereby reducing the heat generation rate while maintaining the fast neutron flux that produces displacement damage similar to fast reactor type conditions. However, rapid insertion and ejection of rabbit capsules containing a strong neutron absorber causes a reactivity response in the reactor that has the potential to engage the HFIR safety response system which could result in an unplanned shutdown. Therefore, a set of tests were performed to provide the data needed to establish limits on the reactivity worth that can be ejected from the hydraulic facility without causing a reactor shutdown. This paper will describe the design, operation, and results of the reactivity measurements undertaken to understand the reactor response to insertion of the gadolinium-lined rabbit capsules. (author)

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

    Science.gov (United States)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N. (U. of Cincinnati)

    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.

  18. Spheromak reactor with poloidal flux-amplifying transformer

    Science.gov (United States)

    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.

  19. Neutron flux measurements in PUSPATI Triga Reactor

    International Nuclear Information System (INIS)

    Neutron flux measurement in the PUSPATI TRIGA Reactor (PTR) was initiated after its commissioning on 28 June 1982. Initial measured thermal neutron flux at the bottom of the rotary specimen rack (rotating) and in-core pneumatic terminus were 3.81E+11 n/cm2 sec and 1.10E+12n/cm2 sec respectively at 100KW. Work to complete the neutron flux data are still going on. The cadmium ratio, thermal and epithermal neutron flux are measured in the reactor core, rotary specimen rack, in-core pneumatic terminus and thermal column. Bare and Cadmium covered gold foils and wires are used for the above measurement. The activities of the irradiated gold foils and wires are determined using Ge(Li) and hyperpure germinium detectors. (author)

  20. Determination flux in the Reactor JEN-1

    International Nuclear Information System (INIS)

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 μ gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs

  1. Initial experiments with the Nevis Cyclotron, the Brookhaven Cosmotron, the Brookhaven AGS and their effects on high energy physics

    International Nuclear Information System (INIS)

    The first experiment at the Nevis Cyclotron by Bernardini, Booth and Lindenbaum demonstrated that nuclear stars are produced by a nucleon-nucleon cascade within the nucleon. This solved a long standing problem in Cosmic rays and made it clear that where they overlap cosmic ray investigation would not be competitive with accelerator investigations. The initial experiments at the Brookhaven Cosmotron by Lindenbaum and Yuan demonstrated that low energy pion nucleon scattering and pion production were unexpectedly mostly due to excitation of the isotopic spin = angular momentum = 3/2 isobaric state of the nucleon. This contradicted the Fermi statistical theory and led to the Isobar model proposed by the author and a collaborator. The initial experiments at the AGS by the author and collaborators demonstrated that the Pomeronchuck Theorem would not come true till at least several hundred GeV. These scattering experiments led to the development of the ''On-line Computer Technique'' by the author and collaborators which is now the almost universal technique in high energy physics. The first accomplishment which flowed from this technique led to contradiction of the Regge pole theory as a dynamical asymptotic theory, by the author and collaborators. The first critical experimental proof of the forward dispersion relation in strong interactions was accomplished by the author and collaborators. They were then used as a crystal ball to predict that ''Asymptopia''---the theoretically promised land where all asymptotic theorems come true---would not be reached till at least 25,000 BeV and probably not before 1,000,000 BeV. 26 refs., 11 figs., 2 tabs

  2. Reactor antineutrino fluxes - status and challenges

    CERN Document Server

    Huber, Patrick

    2016-01-01

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

  3. Fast neutron flux in heavy water reactors

    International Nuclear Information System (INIS)

    The possibility of calculating the fast neutron flux in a natural uranium-heavy water lattice by superposition of the individual contributions of the different fuel elements was verified using a one-dimension Monte-Carlo code. The results obtained are in good agreement with experimental measurements done in the core and reflector of the reactor AQUILON. (author)

  4. HFIR [High-Flux Isotope Reactor] irradiation facilities improvements: Completion of the HIFI [High Irradiation Facilities Improvements] project

    International Nuclear Information System (INIS)

    The HFIR Irradiation Facilities Improvements (HIFI) Project has now been completed. In August 1986, Phase I of the project was completed, providing the capability to perform instrumented irradiation experiments in the target region of the HFIR. In June 1987, Phase II of the project was completed with the assembly in the reactor mockup of all the components necessary to operate up to eight 46-mm-diam instrumented experiments in the removable beryllium region of the HFIR. In conjuntion with the installation of Phase I components, the first instrumented target capsule was installed to determine more accurately the probable temperature in the uninstrumented target capsules previously irradiated as part of the Japan/US fusion materials program. Data from this experiment indicate close agreement with expected temperatures in all positions except those at the extreme ends of the capsule. These data provide a more accurate axial gamma heating rate profile that will allow for better design of future HFIR target irradiation capsules

  5. Optimization of steady-state beam-driven tokamak reactors

    International Nuclear Information System (INIS)

    Recent developments in neutral beam technology prompt us to reconsider the prospects for steady-state tokamak reactors. A mathematical reactor model is developed that includes the physics of beam-driven currents and reactor power balance, as well as reactor and beam system costs. This model is used to find the plasma temperatures that minimize the reactor cost per unit of net electrical output. The optimum plasma temperatures are nearly independent of β and are roughly twice as high as the optimum temperatures for ignited reactors. If beams of neutral deuterium atoms with near-optimum energies of 1 to 2 MeV are used to drive the current in a reactor the size of the International Tokamak Reactor, then the optimum temperatures are typically T /SUB e/ approx. = 12 to 15 keV and T /SUB i/ approx. = 17 to 21 keV for a wide range of model parameters. Net electrical output rises rapidly with increasing deuterium beam energy for E /SUB b/ less than or equal to 400 keV, but rises only slowly above E /SUB b/ about 1 MeV. We estimate that beam-driven steady-state reactors could be economically competitive with pulsed-ignition reactors if cyclic-loading problems limit the toroidal magnetic field strength of pulsed reactors to less than or equal to 85% of that allowed in steady-state reactors

  6. High flux source of cold rubidium atoms

    International Nuclear Information System (INIS)

    We report on the production of a continuous, slow, and cold beam of 87Rb atoms with an extremely high flux of 3.2x1012 atoms/s, a transverse temperature of 3 mK, and a longitudinal temperature of 90 mK. We describe the apparatus created to generate the atom beam. Hot atoms are emitted from a rubidium candlestick atomic beam source and transversely cooled and collimated by a 20 cm long atomic collimator section, boosting overall beam flux by a factor of 50. The Rb atomic beam is then decelerated and longitudinally cooled by a 1 m long Zeeman slower

  7. High flux source of cold rubidium atoms

    Science.gov (United States)

    Slowe, Christopher; Vernac, Laurent; Hau, Lene Vestergaard

    2005-10-01

    We report on the production of a continuous, slow, and cold beam of Rb87 atoms with an extremely high flux of 3.2×1012atoms/s, a transverse temperature of 3mK, and a longitudinal temperature of 90mK. We describe the apparatus created to generate the atom beam. Hot atoms are emitted from a rubidium candlestick atomic beam source and transversely cooled and collimated by a 20cm long atomic collimator section, boosting overall beam flux by a factor of 50. The Rb atomic beam is then decelerated and longitudinally cooled by a 1m long Zeeman slower.

  8. Optimization of steady-state beam-driven tokamak reactors

    International Nuclear Information System (INIS)

    Recent developments in neutral beam technology prompt us to reconsider the prospects for steady-state tokamak reactors. A mathematical reactor model is developed which includes the physics of beam-driven currents and reactor power balance, as well as reactor and beam system costs. This model is used to find the plasma temperatures which minimize the reactor cost per unit of net electrical output. The optimum plasma temperatures are nearly independent of β and are roughly twice as high as the optimum temperatures for ignited reactors. If beams of neutral deuterium atoms with near-optimum energies of 1 to 2 MeV are used to drive the current in an INTOR-sized reactor, then the optimum temperatures are typically T/sub e/ approx. = 12 to 15 keV and T/sub i/ approx. = keV for a wide range of model parameters. Net electrical output rises rapidly with increasing deuterium beam energy for E/sub b/ less than or equal to 400 keV, but rises only slowly above E/sub b/ approx. 1 MeV. We estimate that beam-driven steady-state reactors could be economically competitive with pulsed-ignition reactors if cyclic-loading problems limit the toroidal magnetic field strength of pulsed reactors to less than or equal to 85% of that allowed in steady-state reactors

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

    International Nuclear Information System (INIS)

    On the title date, at ORNL, area radiation alarms sounded during a routine transfer of a shielding cask (containing 60 Ci192Ir) 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Karasiov, A.V. [D.V. Efremov Scientific Rresearch Institute of Electrophysical Apparatus, St. Petersburg (Russian Federation); Greenwood, L.R. [Pacific Northwest Laboratory, Richland, WA (United States)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-06-26

    The objective of the FeCrAl embrittlement screening tests being conducted through the use of Oak Ridge National Laboratories (ORNL) High Flux Isotope Reactor is to provide data on the radiation-induced changes in the mechanical properties including radiation-induced hardening and embrittlement through systematic testing and analysis. Data developed on the mechanical properties will be supported by extensive microstructural evaluations to assist in the development of structure-property relationships and provide a sound, fundamental understanding of the performance of FeCrAl alloys in intense neutron radiation fields. Data and analysis developed as part of this effort will be used to assist in the determination of FeCrAl alloys as a viable material for commercial light water reactor (LWR) applications with a primary focus as an accident tolerant cladding.

  13. Feasibility of a laser or charged-particle-beam fusion-reactor concept with direct electric generation by magnetic-flux compression

    International Nuclear Information System (INIS)

    A new concept for an inertial-confinement fusion reactor is described which, because of its fundamentally different approach to blanket geometry and energy conversion, makes possible a unique combination of high efficiency, high power density, and low radioactivity. The conventional blanket is replaced with a liquid-density mass of lithium contiguously surrounding the fusion yield. This compact blanket configuration produces the maximum shock-induced kinetic energy in liquid metal and the maximum neutron absorption per unit mass. The shock-induced kinetic energy of the liquid lithium is converted directly to electricity with high efficiency by work done against a pulsed normal-conducting magnetic field applied to the exterior of the lithium

  14. Brookhaven highlights - Brookhaven National Laboratory 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    This report highlights research conducted at Brookhaven National Laboratory in the following areas: alternating gradient synchrotron; physics; biology; national synchrotron light source; department of applied science; medical; chemistry; department of advanced technology; reactor; safety and environmental protection; instrumentation; and computing and communications.

  15. The SINQ project. High flux without enrichment problems?

    International Nuclear Information System (INIS)

    A spallation neutron source (SINQ) designed for operation in a continuous mode is presently under construction at The Paul Scherrer Institute in Switzerland and is scheduled for completion in 1994. The waste beam from an isochronous proton cyclotron which is being upgraded to deliver around 1.5 mA of 590 MeV protons will be used after passing through two targets for meson production. The neutron flux in the D2O moderator surrounding the Pb-Bi spallation target is anticipated to amount to some 1014 n/cm2s per mA of beam current on target. This will put SINQ in the regime of most of the medium flux reactors which are presently being considered for operation with low fuel enrichment. Since SINQ will be the world's first spallation neutron source of that design and power level it is difficult to predict, how far the technology can be taken to build spallation neutron sources whose time average neutron flux would reach to level of advanced high flux reactors now possible with highly enriched uranium as fuel material. Another uncertainty factor is the reliability and operational stability of an accelerator needed to achieve such a high power source, which also has never been built so far. Taking everything together, it must be stated that at the very minimum a substantial development effort in a number of fields would be required, most likely leading over various intermediate stages before spallation neutron sources can compete in flux level with advanced high flux reactor designs. (orig.)

  16. Boron neutron capture therapy of malignant brain tumors at the Brookhaven Medical Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joel, D.D.; Coderre, J.A.; Chanana, A.D. [Brookhaven National Lab., Upton, NY (United States). Medical Dept.

    1996-12-31

    Boron neutron capture therapy (BNCT) is a bimodal form of radiation therapy for cancer. The first component of this treatment is the preferential localization of the stable isotope {sup 10}B in tumor cells by targeting with boronated compounds. The tumor and surrounding tissue is then irradiated with a neutron beam resulting in thermal neutron/{sup 10}B reactions ({sup 10}B(n,{alpha}){sup 7}Li) resulting in the production of localized high LET radiation from alpha and {sup 7}Li particles. These products of the neutron capture reaction are very damaging to cells, but of short range so that the majority of the ionizing energy released is microscopically confined to the vicinity of the boron-containing compound. In principal it should be possible with BNCT to selectively destroy small nests or even single cancer cells located within normal tissue. It follows that the major improvements in this form of radiation therapy are going to come largely from the development of boron compounds with greater tumor selectivity, although there will certainly be advances made in neutron beam quality as well as the possible development of alternative sources of neutron beams, particularly accelerator-based epithermal neutron beams.

  17. High-Flux Femtosecond X-Ray Emission from Controlled Generation of Annular Electron Beams in a Laser Wakefield Accelerator.

    Science.gov (United States)

    Zhao, T Z; Behm, K; Dong, C F; Davoine, X; Kalmykov, S Y; Petrov, V; Chvykov, V; Cummings, P; Hou, B; Maksimchuk, A; Nees, J A; Yanovsky, V; Thomas, A G R; Krushelnick, K

    2016-08-26

    Annular quasimonoenergetic electron beams with a mean energy in the range 200-400 MeV and charge on the order of several picocoulombs were generated in a laser wakefield accelerator and subsequently accelerated using a plasma afterburner in a two-stage gas cell. Generation of these beams is associated with injection occurring on the density down ramp between the stages. This well-localized injection produces a bunch of electrons performing coherent betatron oscillations in the wakefield, resulting in a significant increase in the x-ray yield. Annular electron distributions are detected in 40% of shots under optimal conditions. Simultaneous control of the pulse duration and frequency chirp enables optimization of both the energy and the energy spread of the annular beam and boosts the radiant energy per unit charge by almost an order of magnitude. These well-defined annular distributions of electrons are a promising source of high-brightness laser plasma-based x rays. PMID:27610860

  18. Brookhaven highlights, October 1979-September 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    Highlights are given for the research areas of the Brookhaven National Laboratory. These areas include high energy physics, physics and chemistry, life sciences, applied energy science (energy and environment, and nuclear energy), and support activities (including mathematics, instrumentation, reactors, and safety). (GHT)

  19. Brookhaven highlights, October 1979-September 1980

    International Nuclear Information System (INIS)

    Highlights are given for the research areas of the Brookhaven National Laboratory. These areas include high energy physics, physics and chemistry, life sciences, applied energy science (energy and environment, and nuclear energy), and support activities (including mathematics, instrumentation, reactors, and safety)

  20. Neutron beam facilities at the Australian Replacement Research Reactor

    International Nuclear Information System (INIS)

    Australia is building a research reactor to replace the HIFAR reactor at Lucas Heights by the end of 2005. Like HIFAR, the Replacement Research Reactor will be multipurpose with capabilities for both neutron beam research and radioisotope production. It will be a pool-type reactor with thermal neutron flux (unperturbed) of 4 x 1014 n/cm2/sec and a liquid D2 cold neutron source. Cold and thermal neutron beams for neutron beam research will be provided at the reactor face and in a large neutron guide hall. Supermirror neutron guides will transport cold and thermal neutrons to the guide hall. The reactor and the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP S.E. under contract. The neutron beam instruments will be developed by ANSTO, in consultation with the Australian user community. This status report includes a review the planned scientific capabilities, a description of the facility and a summary of progress to date. (author)

  1. Australia's new high performance research reactor

    International Nuclear Information System (INIS)

    A contract for the design and construction of the Replacement Research Reactor was signed in July 2000 between ANSTO and INVAP from Argentina. Since then the detailed design has been completed, a construction authorization has been obtained, and construction has commenced. The reactor design embodies modern safety thinking together with innovative solutions to ensure a highly safe and reliable plant. Also significant effort has been placed on providing the facility with diverse and ample facilities to maximize its use for irradiating material for radioisotope production as well as providing high neutron fluxes for neutron beam research. The project management organization and planing is commensurate with the complexity of the project and the number of players involved. (author)

  2. The Phase I/II BNCT Trials at the Brookhaven medical research reactor: Critical considerations

    International Nuclear Information System (INIS)

    A phase I/II clinical trial of boronophenylalanine-fructose (BPA-F) mediated boron neutron capture therapy (BNCT) for Glioblastoma Multiforme (GBM) was initiated at Brookhaven National Laboratory (BNL) in 1994. Many critical issues were considered during the design of the first of many sequential dose escalation protocols. These critical issues included patient selection criteria, boron delivery agent, dose limits to the normal brain, dose escalation schemes for both neutron exposure and boron dose, and fractionation. As the clinical protocols progressed and evaluation of the tolerance of the central nervous system (CNS) to BPA-mediated BNCT at the BMRR continued new specifications were adopted. Clinical data reflecting the progression of the protocols will be presented to illustrate the steps taken and the reasons behind their adoption. (author)

  3. The NASA Space Radiation Laboratory at Brookhaven National Laboratory: Preparation and delivery of ion beams for space radiation research

    International Nuclear Information System (INIS)

    The NASA Space Radiation Laboratory (NSRL) at Brookhaven National Laboratory (BNL) was commissioned in October 2002 and became operational in July 2003. The NSRL was constructed in collaboration with NASA for the purpose of performing space radiation research as part of the NASA space program. The NSRL can accept a wide variety of ions from BNL's Collider Accelerator Department (CAD) Booster accelerator. These ion beams are extracted from the accelerator with kinetic energies ranging from 0.05 to 3 GeV/nucleon. Many different beam conditions have been produced for experiments at NSRL. The facilities at BNL and the design of the NSRL facility permit a wide variety of beams to be produced with a great degree of flexibility in the delivery of ion beams to experiments. In this report we will describe the facility and its performance over the eight experimental run periods that have taken place since it became operational. We will also describe the current and future capabilities of the NSRL.

  4. Increasing the neutron flux at the beam tube position of the FRG-1

    International Nuclear Information System (INIS)

    The GKSS research center Geesthacht GmbH operates the MTR-type swimming pool research reactor FRG-1 (5MW) for 39 years. The FRG-1 has been converted in February 1991 from HEU (93 percent) to LEU (20 percent) in one step and at that time the core size was reduced from 49 to 26 fuel elements. Consequently the thermal neutron flux in beam tube positions could be increased by more than a factor of two. It is the strong intention of GKSS to continue the operation of the FRG-1 research reactor for at least an additional 15 years with high availability and utilization. The reactor has been operated during 1996 for more than 240 full power days. To prepare the FRG-1 for an efficient future use, a large set of nuclear calculation have been performed to reduce the core size in a second step step from the current 26 fuel elements to 12 fuel elements. To achieve this reduction the fuel loading has to be increased from 3,7 g U/cc to 4,8 g U/cc. The calculational results indicate that the increase in thermal neutron flux for the beam tube is between 50 percent and 160 percent depending on the position of the beam tube. The maximum axial integrated thermal neutron flux will be at the position of the cold neutron source increased 7.5 x 1013 to 12 X 1013 n/cm2 sec. The constructive modification for the new core facilities (gride plate with fraud and the frame for the reactor core) are finished and the application for the license is on the way. The thermohydraulic and safety calculations are on the way and will be completed autumn of 1998 to allow the application for a license for the core size reduction end of 1998. Comparing this conversion procedure with the first conversion procedure we are hopefully looking for a license in 2000. (author)

  5. Neutron spectra at two beam ports of a TRIGA Mark III reactor loaded with HEU fuel.

    Science.gov (United States)

    Vega-Carrillo, H R; Hernández-Dávila, V M; Aguilar, F; Paredes, L; Rivera, T

    2014-01-01

    The neutron spectra have been measured in two beam ports, one radial and another tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research in Mexico. Measurements were carried out with the reactor core loaded with high enriched uranium fuel. Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a (6)LiI(Eu) scintillator and 2, 3, 5, 8, 10 and 12 in.-diameter high-density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code. For each spectrum total flux, mean energy and ambient dose equivalent were determined. Measured spectra show fission, epithermal and thermal neutrons, being harder in the radial beam port. PMID:23746708

  6. Design of neutron beams for boron neutron capture therapy in a fast reactor

    International Nuclear Information System (INIS)

    The BNCT (Boron Neutron Capture Therapy) technique makes use of thermal or epithermal neutrons to irradiate tumours previously loaded with 10B. Reactors are currently seen as a suitable neutron source for BNCT implementation, due to the high intensity of the flux they can provide. The TAPIRO reactor, that is located at the ENEA Casaccia Centre near Rome, is a low-power fast-flux research reactor that can be usefully employed for this application. In this work computer simulations were carried out on this reactor to obtain epithermal and thermal neutron beams for the application of BNCT in Italy in the framework of a specific research program. Comparisons with measurements are also reported. Using the MCNP-4B code, Monte Carlo calculations were carried out to determine the materials suitable for the design of the thermal and epithermal columns. Various arrangements of reflector and moderator materials have been investigated to achieve the desired experimental constraints. On the basis of these calculations, a thermal column was designed and installed in the TAPIRO reactor to perform preliminary experiments on small laboratory animals. For the planning of a therapy treatment of gliomas on larger size animals, several material configurations were investigated in the search for an optimal epithermal facility. The aim of the present study is to indicate how a fast research reactor can be successfully modified for generating neutron beams suitable for BNCT applications. (author)

  7. Investigation of the delay in pressure vessel embrittlement specimen analysis for the Oak Ridge National Laboratory High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Analysis of the investigative data pertaining to this incident reveals the following conditions as key findings and probable causes: (1) The contractor failed to properly implement the surveillance program for monitoring reactor pressure vessel embrittlement. (2) Contractor and DOE organizations provided less than adequate oversight and independent overview, especially by not requiring operating organizations to provide documented evidence to substantiate claims that there was ''no problem'' with respect to embrittlement. (3) Although the temperature limitation for reactor pressurization identified in the Technical Specifications was never violated, the basis of this safety limitation was violated. (4) The basis for concluding that there would be no embrittlement of the pressure vessel steel over the expected life of the reactor is questionable. (5) The contractor and DOE failed to make the surveillance program visible by incorporating it in the Technical Specifications. (6) The Accident Analysis/Final Safety Analysis Report was never adequately reviewed and updated subsequent to its initial issuance. (7) Surveillance specimen analysis was incomplete and never transmitted to reactor operating personnel in a usable format prior to November 1986. (8) There was extensive delays (many years) in the testing, analysis, and reporting of surveillance program results

  8. Measurement of adjoint flux at the RB reactor

    International Nuclear Information System (INIS)

    The adjoint flux is of the great importance for determination of kinetic parameters of nuclear reactor (ρ, l and βeff) and for the interpretation of experiments with reactivity perturbations. In experimental reactor physics there are a few methods for the adjoint flux measurements. The method of reactivity perturbations with adequate samples is used for thermal reactors. According to the theory of reactivity perturbations the reactivity change due to sample of thermal neutrons absorbing material is proportional to product of flux and adjoint flux of thermal neutrons (Φ2(r)Φ=2(r)). The reactivity change due to fissionable nuclide is proportional to product of thermal neutron flux and adjoint flux of fast neutrons (Φ2(r)Φ=2(r)). The axial distribution of adjoint flux is determined by reactivity measurements and measurements of axial distribution of thermal neutron flux. Thi results of this measurement will be used for interpretation of other experiments with reactivity perturbations at the RB reactor

  9. Production capabilities in US nuclear reactors for medical radioisotopes

    Energy Technology Data Exchange (ETDEWEB)

    Mirzadeh, S.; Callahan, A.P.; Knapp, F.F. Jr. [Oak Ridge National Lab., TN (United States); Schenter, R.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1992-11-01

    The availability of reactor-produced radioisotopes in the United States for use in medical research and nuclear medicine has traditionally depended on facilities which are an integral part of the US national laboratories and a few reactors at universities. One exception is the reactor in Sterling Forest, New York, originally operated as part of the Cintichem (Union Carbide) system, which is currently in the process of permanent shutdown. Since there are no industry-run reactors in the US, the national laboratories and universities thus play a critical role in providing reactor-produced radioisotopes for medical research and clinical use. The goal of this survey is to provide a comprehensive summary of these production capabilities. With the temporary shutdown of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) in November 1986, the radioisotopes required for DOE-supported radionuclide generators were made available at the Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR). In March 1988, however, the HFBR was temporarily shut down which forced investigators to look at other reactors for production of the radioisotopes. During this period the Missouri University Research Reactor (MURR) played an important role in providing these services. The HFIR resumed routine operation in July 1990 at 85 MW power, and the HFBR resumed operation in June 1991, at 30 MW power. At the time of the HFBR shutdown, there was no available comprehensive overview which could provide information on status of the reactors operating in the US and their capabilities for radioisotope production. The obvious need for a useful overview was thus the impetus for preparing this survey, which would provide an up-to-date summary of those reactors available in the US at both the DOE-funded national laboratories and at US universities where service irradiations are currently or expected to be conducted.

  10. Production capabilities in US nuclear reactors for medical radioisotopes

    International Nuclear Information System (INIS)

    The availability of reactor-produced radioisotopes in the United States for use in medical research and nuclear medicine has traditionally depended on facilities which are an integral part of the US national laboratories and a few reactors at universities. One exception is the reactor in Sterling Forest, New York, originally operated as part of the Cintichem (Union Carbide) system, which is currently in the process of permanent shutdown. Since there are no industry-run reactors in the US, the national laboratories and universities thus play a critical role in providing reactor-produced radioisotopes for medical research and clinical use. The goal of this survey is to provide a comprehensive summary of these production capabilities. With the temporary shutdown of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) in November 1986, the radioisotopes required for DOE-supported radionuclide generators were made available at the Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR). In March 1988, however, the HFBR was temporarily shut down which forced investigators to look at other reactors for production of the radioisotopes. During this period the Missouri University Research Reactor (MURR) played an important role in providing these services. The HFIR resumed routine operation in July 1990 at 85 MW power, and the HFBR resumed operation in June 1991, at 30 MW power. At the time of the HFBR shutdown, there was no available comprehensive overview which could provide information on status of the reactors operating in the US and their capabilities for radioisotope production. The obvious need for a useful overview was thus the impetus for preparing this survey, which would provide an up-to-date summary of those reactors available in the US at both the DOE-funded national laboratories and at US universities where service irradiations are currently or expected to be conducted

  11. High power 1 MeV neutral beam system and its application plan for the international tokamak experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hemsworth, R.S. [ITER Joint Central Team, Naka, Ibaraki (Japan)

    1997-03-01

    This paper describes the Neutral Beam Injection system which is presently being designed for the International Tokamak Experimental Reactor, ITER, in Europe Japan and Russia, with co-ordination by the Joint Central Team of ITER at Naka, Japan. The proposed system consists of three negative ion based neutral injectors, delivering a total of 50 MW of 1 MeV D{sup 0} to the ITER plasma for a pulse length of >1000 s. Each injectors uses a single caesiated volume arc discharge negative ion source, and a multi-grid, multi-aperture accelerator, to produce about 40 A of 1 MeV D{sup -}. This will be neutralized by collisions with D{sub 2} in a sub-divided gas neutralizer, which has a conversion efficiency of about 60%. The charged fraction of the beam emerging from the neutralizer is dumped in an electrostatic residual ion dump. A water cooled calorimeter can be moved into the beam path to intercept the neutral beam, allowing commissioning of the injector independent of ITER. ITER is scheduled to produce its first plasma at the beginning of 2008, and the planning of the R and D, construction and installation foresees the neutral injection system being available from the start of ITER operations. (author)

  12. High temperature reactors

    International Nuclear Information System (INIS)

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

  13. Charge neutralized low energy beam transport at Brookhaven 200 MeV linac

    International Nuclear Information System (INIS)

    The H− magnetron source provides about 100 mA H− beam to be match into the radio-frequency quadrupole accelerator. As H− beam traverses through low energy transport, it ionizes the residual gas and electrons are repelled and positive ions are trapped in the beam, due to negative potential of the beam, providing charge neutralization for the H− beam. The neutralization time for the critical density depends upon the background gas and its pressure. Critical density for xenon gas at 35 keV is about 43 times smaller than that of hydrogen and stripping cross section is only 5 times than that of hydrogen gas. We are using xenon gas to reduce neutralization time and to improve transmission through the 200 MeV linac. We are also using pulse nitrogen gas to improve transmission and stability of polarized H− beam from optically pumped polarized ion source

  14. Charge neutralized low energy beam transport at Brookhaven 200 MeV linac.

    Science.gov (United States)

    Raparia, D; Alessi, J; Atoian, G; Zelenski, A

    2016-02-01

    The H(-) magnetron source provides about 100 mA H(-) beam to be match into the radio-frequency quadrupole accelerator. As H(-) beam traverses through low energy transport, it ionizes the residual gas and electrons are repelled and positive ions are trapped in the beam, due to negative potential of the beam, providing charge neutralization for the H(-) beam. The neutralization time for the critical density depends upon the background gas and its pressure. Critical density for xenon gas at 35 keV is about 43 times smaller than that of hydrogen and stripping cross section is only 5 times than that of hydrogen gas. We are using xenon gas to reduce neutralization time and to improve transmission through the 200 MeV linac. We are also using pulse nitrogen gas to improve transmission and stability of polarized H(-) beam from optically pumped polarized ion source. PMID:26932107

  15. Analytical study on flux distribution in 5 MW HEU [high enriched uranium] and LEU [low enriched uranium] TRR [Teheran Research Reactor] core

    International Nuclear Information System (INIS)

    In HEU to LEU fuel conversion LEU core suffers unformidable changes in core arrangement and fuel element design structure. These lead to some redesign calculations as regard to heat removal and control potentiality. In this paper, some results related to flux distribution are given. Two core configurations with 18 (flat) plates/FE and two types of control elements, oval (8 pl/FE) and Fork-type (12 pl/FE) fork-type were considered. In oval type control element thermal flux depression in LEU fuel as compared to HEU fuel is about 30%. In case of LEU fuel flux distributions are mainly cosine and general power distribution follows more or less the flux shape. Two shuffling patterns were studied and indicate some slight changes in flux level. Our calculations showed that based on reactor operational requirements to an optimum fuel loading and fuel element design characteristics can be reached. (Author)

  16. A high-speed data acquisition system to measure low-level current from self-powered flux detectors in CANDU nuclear reactors

    International Nuclear Information System (INIS)

    Self-powered flux detectors are used in CANDU nuclear power reactors to determine the spatial neutron flux distribution in the reactor core for use by both the reactor control and safety systems. To establish the dynamic response of different types of flux detectors, the Chalk River Nuclear Laboratories have an ongoing experimental irradiation program in the NRU research reactor for which a data acquistion system has been developed. The system described in this paper is used to measure the currents from the detectors both at a slow, regular logging interval, and at a rapid, adaptive rate following a reactor shutdown. Currents that range from 100 pA to 1 mA full scale can be measured from up to 38 detectors and stored at sampling rates of up to 20 samples per second. The dynamic characteristics of the detectors can be computed from the stored records. The data acquisition system comprises a DEC LSI-11/23 microcomputer, dual cartridge disks, floppy disks, a hard copy and a video display terminal. The RT-11 operating system is used and all application programs are written in FORTRAN

  17. Brookhaven segment interconnect

    International Nuclear Information System (INIS)

    We have performed a high energy physics experiment using a multisegment Brookhaven FASTBUS system. The system was composed of three crate segments and two cable segments. We discuss the segment interconnect module which permits communication between the various segments

  18. Bunch-to-bucket injection of linac beam into the Brookhaven AGS

    International Nuclear Information System (INIS)

    A new fast beam chopper has been used to study injection and capture in the AGS. The chopper is a fast beam switch with 10 ns rise and fall times that can be programmed on a bunch-by-bunch basis and is synchronized to the net accelerating voltage of the synchrotron, thus allowing bunch-to-bucket injection of the 200 MeV H- linac beam. The studies so far have concentrated on simple injection scenarios, at reduced intensity, where longitudinal effects are well separated from transverse. The evolution of the pre-bunched beam during the transition from injection to acceleration has been examined. Results have shown the importance of the detailed linac beam energy distribution. The ability to control the longitudinal emittance of the beam with the fast chopper has been used in other machine studies. This report includes a description of a measurement of the longitudinal coupling impedance of the AGS by the beam transfer function technique which utilized the control of longitudinal emittance provided by bunch-to-bucket injection. Plans for improvements to the chopper equipment are also describe. 6 refs., 4 figs

  19. Technological research on Recycling of Actinides and fission products (RAS). Irradiations in the High Flux Reactor (HFR), Petten, Netherlands

    International Nuclear Information System (INIS)

    The purpose of the title irradiations is to study the efficiency and technical feasibility of possible transmutation processes for those long-lived actinides and fission products, that contribute to long-term radiotoxicity and leaking risks of geological storage. A cooperative research program (EFFTRA or Experimental Feasibility of Targets for TRAnsmutation) has been set up for irradiations of technetium, iodine and americium in the thermal reactor HFR and the fast reactor Phenix. A radiation program for fission products is in progress in the HFR. An inert matrix concept is developed, in which the actinide is mixed with a ceramic material, which hardly reacts with neutrons and actinides and containment materials. Irradiation experiments with candidate inert matrices will be carried out in the HFR. Also, the feasibility of transmutation of americium in a thermal spectrum will be demonstrated by means of a long-range experiment in the HFR. Plans are elaborated for the irradiation of plutonium in inert matrices in the HFR to realize an efficient transmutation of existing supplies, both military and civil, of plutonium. 8 figs., 4 tabs., 18 refs

  20. Charge neutralized low energy beam transport at Brookhaven 200 MeV linac

    Energy Technology Data Exchange (ETDEWEB)

    Raparia, D., E-mail: raparia@bnl.gov; Alessi, J.; Atoian, G.; Zelenski, A. [Brookhaven National Laboratory, Upton, New York 11786 (United States)

    2016-02-15

    The H{sup −} magnetron source provides about 100 mA H{sup −} beam to be match into the radio-frequency quadrupole accelerator. As H{sup −} beam traverses through low energy transport, it ionizes the residual gas and electrons are repelled and positive ions are trapped in the beam, due to negative potential of the beam, providing charge neutralization for the H{sup −} beam. The neutralization time for the critical density depends upon the background gas and its pressure. Critical density for xenon gas at 35 keV is about 43 times smaller than that of hydrogen and stripping cross section is only 5 times than that of hydrogen gas. We are using xenon gas to reduce neutralization time and to improve transmission through the 200 MeV linac. We are also using pulse nitrogen gas to improve transmission and stability of polarized H{sup −} beam from optically pumped polarized ion source.

  1. Colliding beam fusion reactor space propulsion system

    Science.gov (United States)

    Wessel, Frank J.; Binderbauer, Michl W.; Rostoker, Norman; Rahman, Hafiz Ur; O'Toole, Joseph

    2000-01-01

    We describe a space propulsion system based on the Colliding Beam Fusion Reactor (CBFR). The CBFR is a high-beta, field-reversed, magnetic configuration with ion energies in the range of hundreds of keV. Repetitively-pulsed ion beams sustain the plasma distribution and provide current drive. The confinement physics is based on the Vlasov-Maxwell equation, including a Fokker Planck collision operator and all sources and sinks for energy and particle flow. The mean azimuthal velocities and temperatures of the fuel ion species are equal and the plasma current is unneutralized by the electrons. The resulting distribution functions are thermal in a moving frame of reference. The ion gyro-orbit radius is comparable to the dimensions of the confinement system, hence classical transport of the particles and energy is expected and the device is scaleable. We have analyzed the design over a range of 106-109 Watts of output power (0.15-150 Newtons thrust) with a specific impulse of, Isp~106 sec. A 50 MW propulsion system might involve the following parameters: 4-meters diameter×10-meters length, magnetic field ~7 Tesla, ion beam current ~10 A, and fuels of either D-He3,P-B11,P-Li6,D-Li6, etc. .

  2. Long Distance Reactor Antineutrino Flux Monitoring

    Science.gov (United States)

    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.

  3. LWR fuel rod testing facilities in high flux reactor (HFT) Petten for investigation of power cycling and ramping behaviour

    International Nuclear Information System (INIS)

    LWR fuel rod irradiation experiments are being performed in HFR's Pool Side Facility (PSF) by means of pressurized boiling water capsules (BWFC). For more than 6 years the major application of these devices has been in performing irradiation programs to investigate the power ramp behaviour of PWR and BWR rods which have been pre-irradiated in power reactors. Irradiation devices with various types of monitoring instrumentation are available, e.g. for fuel rod length, fuel stack length, fuel rod internal pressure and fuel rod central temperature measurements. The application scope covers PWR and BWR fuel rods, with burn-up levels up to 45 MWd/kg(U), max. linear heat generation rates up to 700 W/cm and simulation of constant power change rates between 0.05 and 1000 W/cm.min. The paper describes the different designs of LWR fuel rod testing facilities and associated non-destructive testing devices in use at the HFR Petten. It also addresses the new test capabilities that will become available after exchange of the HFR vessel in 1983. Furthermore it shows some typical results. (author)

  4. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Absolute measurement of neutron fluxes inside the reactor core

    International Nuclear Information System (INIS)

    The subject of this work is the development and study of two methods of neutron measurements in nuclear reactors, the new method of high neutron flux measurements and the Li6-semiconductor neutron spectrometer. This work is presented in four sections: Section I. The introduction explains the need for neutron measurements in reactors. A critical survey is given of the existing methods of high neutron flux measurement and methods of fast neutron spectrum determination. Section II. Theoretical basis of the work of semiconductor counters and their most important characteristics are given. Section III. The main point of this section is in presenting the basis of the new method which the author developed, i.e., the long-tube method, and the results obtained by it, with particular emphasis on absolute measurement of high neutron fluxes. Advantages and limitations of this method are discussed in details at the end of this section. Section IV. A comparison of the existing semiconductor neutron spectrometers is made and their advantages and shortcomings underlined. A critical analysis of the obtained results with the Li6-semiconductor spectrometer with plane geometry is given. A new type of Li6-semiconductor spectrometer is described, its characteristics experimentally determined, and a comparison of it with a classical Li6-spectrometer made (author)

  6. Modification of the radial beam port of ITU TRIGA Mark II research reactor for BNCT applications

    International Nuclear Information System (INIS)

    This paper aims to describe the modification of the radial beam port of ITU (İstanbul Technical University) TRIGA Mark II research reactor for BNCT applications. Radial beam port is modified with Polyethylene and Cerrobend collimators. Neutron flux values are measured by neutron activation analysis (Au–Cd foils). Experimental results are verified with Monte Carlo results. The results of neutron/photon spectrum, thermal/epithermal neutron flux, fast group photon fluence and change of the neutron fluxes with the beam port length are presented. - Highlights: • Using MCNP5, radial beam port of ITU TRIGA Mark II research reactor is modified. • Polyethylene and Cerrobend collimators are used to modify the beam port. • Results of two-group neutron/photon flux are presented. • Monte Carlo results are compared with experimental results

  7. RF Beam control system for the Brookhaven relativistic heavy ion collider, RHIC

    International Nuclear Information System (INIS)

    The Relativistic Heavy Ion Collider, RHIC, is two counter-rotating rings with six interaction points. The RF Beam Control system for each ring will control two 28 MHz cavities for acceleration, and five 197 MHz cavities for preserving the 5 ns bunch length during 10 hour beam stores. Digital technology is used extensively in: Direct Digital Synthesis of rf signals and Digital Signal Processing for, the realization of state-variable feedback loops, real-time calculation of rf frequency, and bunch-by-bunch phase measurement of the 120 bunches. DSP technology enables programming the parameters of the feedback loops in order to obtain closed-loop dynamics that are independent of synchrotron frequency

  8. At the Cutting Edge of Bright Beams: The NSLS Source Development Lab (432nd Brookhaven Lecture)

    International Nuclear Information System (INIS)

    Inspired by the discoveries with synchrotron light at the National Synchrotron Light Source (NSLS) and similar facilities around the world, researchers are looking for more brilliant beams of light. To develop this next-generation of light sources, accelerator physicists at the NSLS Source Development Laboratory (SDL) make use of a magnesium photocathode irradiated by ultraviolet laser light to produce electron beams of unprecedented brightness. As Murphy will describe in his talk, he and fellow researchers have developed various techniques to catch molecules and atoms in action. In one recent study, the researchers used a laser to control the pulse duration of light from a free-electron laser (FEL), a type of light source with a potential peak brightness up to one billion times higher than that of ordinary synchrotron light. In another technique, Murphy and his colleagues generated extremely short pulses of terahertz radiation that are the highest intensity of their type ever produced.

  9. Neutron flux determination at the IPR-R1 Triga Mark I neutron beam extractor

    International Nuclear Information System (INIS)

    The IPR-R1 Triga Mark I Reactor located at the CDTN/CNEN, Belo Horizonte, Brazil, has been operating since November of 1960. In this work, measurements of thermal and epithermal neutron flux along the IPR-R1 neutron beam extractor were performed by neutron activation of reference materials using the two foils method. The obtained results were compared with results from two previous works: an experimental measurement done in a previous reactor core configuration and a numerical work made by Monte Carlo simulation using the actual reactor core configuration. The main purpose of this work is to update the measured data to the actual reactor core configuration. (author)

  10. Determination of maximum reactor power level consistent with the requirement that flow reversal occurs without fuel damage

    International Nuclear Information System (INIS)

    The High Flux Beam Reactor (HFBR) operated by Brookhaven National Laboratory (BNL) employs forced downflow for heat removal during normal operation. In the event of total loss of forced flow, the reactor will shutdown and the flow reversal valves open. When the downward core flow becomes sufficiently small then the opposing thermal buoyancy induces flow reversal leading to decay heat removal by natural convection. There is some uncertainty as to whether the natural circulation is adequate for decay heat removal after 60 MW operation. BNL- staff carried out a series of calculations to establish the adequacy of flow reversal to remove decay heat. Their calculations are based on a natural convective CHF model. The primary purpose of the present calculations is to review the accuracy and applicability of Fauske's CHF model for the HFBR, and the assumptions and methodology employed by BNL-staff to determine the heat removal limit in the HFBR during a flow reversal and natural convection situation

  11. Neutron flux and fluence determination for BWR reactors

    International Nuclear Information System (INIS)

    Measurements of gamma emission rates from Fe and Cu dosimeters extracted from a BWR type reactor vessel were carried out in order to determine their total activity. The dosimeter's activity is related to the neutron flux there by taking into account the reactor material's embrittlement caused by neutron bombardment. The dosimeters were taken out after the first reactor operation cycle. From gamma radioactivity measurements of these dosimeters, neutron flux and fluence were calculated. These parameters are used in the determination of shift and adjusted reference temperature values needed for the development of pressure-temperature curves used during reactor operation

  12. Neutron flux measurements with Monte Carlo verification at the thermal column of a TRIGA MARK II reactor: Feasibility study for a BNCT facility

    International Nuclear Information System (INIS)

    The treatment of the malignant brain tumor through Boron Neutron Capture Therapy (BNCT) requires a high-flux neutron source. The Malaysian TRIGA Mark II reactor was investigated for a proposed BNCT facility. The neutron flux was measured along the central stringer of the thermal column and the outermost positions of the other stringers. The unfolding foil method was applied here. We have used Al, As, Au, Co, In, Mo, Ni and Re foils and Cd as a cover with 19 useful reactions in this study. The infinitely diluted foil activity was calculated and used in the SAND-II code (Spectrum Analysis by Neutron Detectors) to calculate the neutron flux. The reactor was also simulated using Monte Carlo code (MCNP5) and the neutron flux was calculated along the thermal column. The measured and calculated neutron flux along the thermal column show good agreement. The minimum epithermal neutron intensity required for BNCT is achieved up to position 22 with a mixed neutron-gamma beam. A suggested MCNP simulated modification of the reactor thermal column increased the neutron flux at distant positions from the reactor core but the epithermal neutron part was below the minimum requirement for a BNCT facility. The photon flux calculations along the thermal column show relatively high results which should be filtered. The calculation of the neutron and gamma dose in a head phantom (water) indicated that the available neutron spectrum requires modifications to increase the epithermal part of the neutrons and filter the gamma ray contamination. (author)

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

    International Nuclear Information System (INIS)

    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%

  14. In-pile tritium release behaviour of lithiummetatitanate produced by extrusion-spheroidisation-sintering process in EXOTIC-9/1 in the high flux reactor, Petten

    Energy Technology Data Exchange (ETDEWEB)

    Peeters, M.M.W. [N.R.G., P.O. Box 25, 1755 ZG Petten (Netherlands)], E-mail: peeters@nrg-nl.com; Magielsen, A.J.; Stijkel, M.P.; Laan, J.G. van der [N.R.G., P.O. Box 25, 1755 ZG Petten (Netherlands)

    2007-10-15

    The irradiation programme EXOTIC (extraction of tritium in ceramics) is carried out within the European framework for the development of the helium cooled pebble bed concept. The EXOTIC-9/1 is the latest experiment in the series of EXOTICs that are irradiated in the high flux reactor in Petten. Tritium release and inventory in lithium containing ceramic pebbles are key properties to be tested in a TBM. New production routes of pebbles are developed, leading to different thermomechanical and tritium release properties. The objective of the EXOTIC-9/1 is to study in-pile tritium release behaviour of the latest developed lithiummetatitanate pebbles (Li{sub 2}TiO{sub 3}). The pebbles are produced by a extrusion-spheroidisation-sintering process at CEA. The new pebbles differ with respect to porosity from the lithiummetatitanate ceramics tested in the previous EXOTIC 8 programme. The pebbles have diameter in the range from 0.6 to 0.8 mm. Irradiation of EXOTC-9/1 started at 24 March 2005, and will continue until the end of 2006, in total about 400 irradiation days. The temperature is varied between 340 and 580 deg. C. Begin of Life (BOL) tritium production rate is 0.56 mCi/min. Based upon the in-pile tritium release measurements and the analysis of the tritium residence time it can be concluded that tritium release in the new batch of the high density Li{sub 2}TiO{sub 3} pebbles irradiated in EXOTIC 9/1 is rather slow compared to the ceramics irradiated in the EXOTIC 8 irradiation campaign. In this paper, the in-pile tritium behaviour will be reported during normal operation and during transients in temperature, purge gas chemistry and gasflow. The collected data is compared to tritium release data from ceramics irradiated in previous EXOTIC experiments with respect to tritium inventory, residence time and porosity.

  15. A high-speed beam of lithium droplets for collecting diverted energy and particles in ITER [International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    A high-speed (160m/s) beam (0.14 x 0.86m) of liquid-lithium droplets passing through the divertor region(s) below (and above) the main plasma has the potential to replace and out-perform ''conventional'' solid divertor plates in both heat and particle removal. In addition to superior heat-collection properties, the lithium beam would: remove impurities; require low power to circulate the lithium; exhibit low-recycle divertor operation compatible with lower-hybrid current drive, H-mode plasma confinement, and no flow reversal in the edge plasma; be insensitive to plasma shifts; and finally protect solid structures from the plasma thermal energy for those disruptions that deposit energy preferentially into the divertor while simultaneously being rapidly re-established after a major disruption. Scoping calculations identifying the beam configuration and the droplet dynamics, including formation, MHD effects, gravitational effects, thermal response and hydrodynamics, are presented. Limitations and uncertainties are also discussed. 20 refs., 6 figs., 3 tabs

  16. Modification of the radial beam port of ITU TRIGA Mark II research reactor for BNCT applications.

    Science.gov (United States)

    Akan, Zafer; Türkmen, Mehmet; Çakir, Tahir; Reyhancan, İskender A; Çolak, Üner; Okka, Muhittin; Kiziltaş, Sahip

    2015-05-01

    This paper aims to describe the modification of the radial beam port of ITU (İstanbul Technical University) TRIGA Mark II research reactor for BNCT applications. Radial beam port is modified with Polyethylene and Cerrobend collimators. Neutron flux values are measured by neutron activation analysis (Au-Cd foils). Experimental results are verified with Monte Carlo results. The results of neutron/photon spectrum, thermal/epithermal neutron flux, fast group photon fluence and change of the neutron fluxes with the beam port length are presented. PMID:25746919

  17. Optimization of neutron flux distribution in Isotope Production Reactor

    International Nuclear Information System (INIS)

    In order to optimize the thermal neutrons flux distribution in a Radioisotope Production and Research Reactor, the influence of two reactor parameters was studied, namely theVmod/Vcomb ratio and the core volume. The reactor core is built with uranium oxide pellets (UO2) mounted in rod clusters, with an enrichment level of ∼3 %, similar to LIGHT WATER POWER REATOR (LWR) fuel elements. (author)

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

    Science.gov (United States)

    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

  19. Excitation of neutron flux waves in reactor core transients

    International Nuclear Information System (INIS)

    An analysis of the excitation of neutron flux waves in reactor core transients has been performed. A perturbation theory solution has been developed for the time-dependent thermal diffusion equation in which the absorption cross section undergoes a rapid change, as in a PWR rod ejection accident (REA). In this analysis the unperturbed reactor flux states provide the basis for the spatial representation of the flux solution. Using a simplified space-time representation for the cross section change, the temporal integrations have been carried out and analytic expressions for the modal flux amplitudes determined. The first order modal excitation strength is determined by the spatial overlap between the initial and final flux states, and the cross section perturbation. The flux wave amplitudes are found to be largest for rapid transients involving large reactivity perturbations

  20. Use of neutron beams for low and medium flux research reactors: R and D programmes in materials science. Report of an advisory group meeting held in Vienna, 29 March - 1 April 1993

    International Nuclear Information System (INIS)

    The report is intended to provide guidelines to research reactor owners and operators for promoting and developing neutron beam based research programmes for solid state studies using neutron scattering techniques. It is expected to benefit ongoing facilities and programmes by encouraging use of improved techniques for detection, signal acquisition, signal processing, etc. and new programmes by assisting in the selection of appropriate equipment, instrument design and research plans. Refs, figs and tabs

  1. The Brookhaven National Laboratory Accelerator Test Facility

    International Nuclear Information System (INIS)

    The Brookhaven National Laboratory Accelerator Test Facility comprises a 50 MeV traveling wave electron linear accelerator utilizing a high gradient, photo-excited, raidofrequency electron gun as an injector and an experimental area for study of new acceleration methods or advanced radiation sources using free electron lasers. Early operation of the linear accelerator system including calculated and measured beam parameters are presented together with the experimental program for accelerator physics and free electron laser studies

  2. Reactor Neutrino Flux Uncertainty Suppression on Multiple Detector Experiments

    CERN Document Server

    Cucoanes, Andi; Cabrera, Anatael; Fallot, Muriel; Onillon, Anthony; Obolensky, Michel; Yermia, Frederic

    2015-01-01

    This publication provides a coherent treatment for the reactor neutrino flux uncertainties suppression, specially focussed on the latest $\\theta_{13}$ measurement. The treatment starts with single detector in single reactor site, most relevant for all reactor experiments beyond $\\theta_{13}$. We demonstrate there is no trivial error cancellation, thus the flux systematic error can remain dominant even after the adoption of multi-detector configurations. However, three mechanisms for flux error suppression have been identified and calculated in the context of Double Chooz, Daya Bay and RENO sites. Our analysis computes the error {\\it suppression fraction} using simplified scenarios to maximise relative comparison among experiments. We have validated the only mechanism exploited so far by experiments to improve the precision of the published $\\theta_{13}$. The other two newly identified mechanisms could lead to total error flux cancellation under specific conditions and are expected to have major implications o...

  3. EU Development of High Heat Flux Components

    OpenAIRE

    Linke, J.; Lorenzetto, P.; Majerus, P.; Merola, M.; Pitzer, D.; Rödig, M.

    2005-01-01

    The development of plasma facing components for next step fusion devices in Europe is strongly focused to ITER. Here a wide spectrum of different design options for the divertor target and the first wall have been investigated with tungsten, CFC, and beryllium armor. Electron beam simulation experiments have been used to determine the performance of high heat flux components under ITER specific thermal loads. Beside thermal fatigue loads with power density levels up to 20MWm(-2), off normal e...

  4. Neutron beam experiments using nuclear research reactors: honoring the retirement of professor Bernard W. Wehring -I. 3. A Comparison of Neutron Beams for BNCT

    International Nuclear Information System (INIS)

    This paper evaluates the potential of the Ohio State University (OSU) Research Reactor (OSURR) with a fission convertor plate (FCP) for clinical boron neutron capture therapy (BNCT). The evaluation uses design methods that were developed for the analysis of the OSU design of an accelerator-based neutron source (ABNS) for BNCT (hereafter called the OSU-ABNS); namely, the in-phantom neutron field assessment parameters, the treatment time (T) and the high-LET absorbed-dose to the tumor (DTumor), were calculated using MCNP. The paper compares an FCP epithermal neutron beam, which is based on the OSURR (hereafter called the OSURR-FCP) with the OSU-ABNS. For completeness, the comparison includes an alternative ABNS design, which was taken from the literature (hereafter called the 7LiF-Al2O3 ABNS), and the Brookhaven Medical Research Reactor (BMRR) epithermal neutron beam for BNCT (hereafter called the BMRR-ENB). The OSURR-FCP design consists of the OSURR, a fission plate, and a moderator/filter assembly. These components were modeled in MCNP. The OSURR is a 500-kW pool-type light water-cooled and moderated reactor that is reflected on two sides with graphite and uses a U3Si2-Al dispersion fuel. The fission plate and moderator/filter assembly, which were modeled, were identical to those specified by Liu. The goal of our analysis was not to perfect an FCP and moderator/filter assembly for the OSURR-FCP. Rather, the intent of our analysis was to determine if, using the FCP and moderator/filter assembly designed by Liu, the OSURR, operating at 100% power, could produce a beam of sufficient intensity to treat human patients with BNCT in a reasonable treatment time. T is the total time required for a BNCT treatment, including all treatment fractions. Since the total dose delivered to the tumor is limited by the tolerance of the surrounding normal tissue, T is defined as the time required to escalate the normal tissue RBE-dose to the tolerance of the normal brain. DTumor is the

  5. Design and characterization of 2.45 GHz electron cyclotron resonance plasma source with magnetron magnetic field configuration for high flux of hyperthermal neutral beam

    International Nuclear Information System (INIS)

    A 2.45 GHz electron cyclotron resonance (ECR) source with a magnetron magnetic field configuration was developed to meet the demand of a hyperthermal neutral beam (HNB) flux on a substrate of more than 1x1015 cm-2 s-1 for industrial applications. The parameters of the operating pressure, ion density, electron temperature, and distance between the neutralization plate and the substrate for the HNB source are specified in a theoretical analysis. The electron temperature and the ion density are measured to characterize the ECR HNB source using a Langmuir probe and optical emission spectroscopy. The parameters of the ECR HNB source are in good agreement with the theoretically specified parameters.

  6. AGS Resonant Extraction with High Intensity Beams

    International Nuclear Information System (INIS)

    The Brookhaven AGS third integer resonant extraction system allows the AGS to provide high quality, high intensity 25.5 GeV/c proton beams simultaneously to four target stations and as many as 8 experiments. With the increasing intensities (over 7 x 1013 protons/pulse) and associated longer spill periods (2.4 to 3 seconds long), they continue to run with low losses and high quality low modulation continuous current beams. Learning to extract and transport these higher intensity beams has required a process of careful modeling and experimentation. They have had to learn how to correct for various instabilities and how to better match extraction and the transport lines to the higher emittance beams being accelerated in the AGS. Techniques employed include ''RF'' methods to smooth out momentum distributions and fine structure. They will present results of detailed multi-particle tracking modeling studies which enabled them to develop a clear understanding of beam loss mechanisms in the transport and extraction process. They report on their status, experiences, and the present understanding of the intensity limitations imposed by resonant extraction and transport to fixed target stations

  7. AGS RESONANT EXTRACTION WITH HIGH INTENSITY BEAMS.

    Energy Technology Data Exchange (ETDEWEB)

    AHRENS,L.; BROWN,K.; GLENN,J.W.; ROSER,T.; TSOUPAS,N.; VANASSELT,W.

    1999-03-29

    The Brookhaven AGS third integer resonant extraction system allows the AGS to provide high quality, high intensity 25.5 GeV/c proton beams simultaneously to four target stations and as many as 8 experiments. With the increasing intensities (over 7 x 10{sup 13} protons/pulse) and associated longer spill periods (2.4 to 3 seconds long), we continue to run with low losses and high quality low modulation continuous current beams.[1] Learning to extract and transport these higher intensity beams has required a process of careful modeling and experimentation. We have had to learn how to correct for various instabilities and how to better match extraction and the transport lines to the higher emittance beams being accelerated in the AGS. Techniques employed include ''RF'' methods to smooth out momentum distributions and fine structure. We will present results of detailed multi-particle tracking modeling studies which enabled us to develop a clear understanding of beam loss mechanisms in the transport and extraction process. We will report on our status, experiences, and the present understanding of the intensity limitations imposed by resonant extraction and transport to fixed target stations.

  8. High-energy gamma-ray beams from Compton-backscattered laser light

    Energy Technology Data Exchange (ETDEWEB)

    Sandorfi, A.M.; LeVine, M.J.; Thorn, C.E.; Giordano, G.; Matone, G.

    1983-01-01

    Collisions of light photons with relativistic electrons have previously been used to produce polarized ..gamma..-ray beams with modest (-10%) resolution but relatively low intensity. In contrast, the LEGS project (Laser + Electron Gamma Source) at Brookhaven will produce a very high flux (>2 x 10/sup 7/ s/sup -1/) of background-free polarized ..gamma.. rays whose energy will be determined to a high accuracy (..delta..E = 2.3 MeV). Initially, 300(420)-MeV ..gamma.. rays will be produced by backscattering uv light from the new 2.5(3.0)-GeV X-ray storage ring of the National Synchrotron Light Source (NSLS). The LEGS facility will operate as one of many passive users of the NSLS. In a later stage of the project, a Free Electron Laser is expectred to extend the ..gamma..-ray energy up to 700 MeV.

  9. Epithermal BNCT neutron beam design for a TRIGA II reactor

    International Nuclear Information System (INIS)

    In Finland a collaborative effort by Helsinki University Central Hospital, MAP Medical Technologies Inc. and VTT Reactor Laboratory has started aiming at BNCT of glioma patients. For this the capabilities of the FiR-1 TRIGA II 250 kW research reactor have been evaluated. The FiR-1 is located in the middle of the Otaniemi campus eight kilometers from the center of Helsinki and four kilometers from the Central Hospital. The power of the reactor was increased in 1965 to 250 kW and the instrumentation modernised in 1981. It is a pool reactor with graphite reflector and a core loading of 3 kg 20w% 235U in the special TRIGA uranium-zirconium hydride fuel (8-12 w% U, 91% Zr, 1% H). The advantages of using a TRIGA reactor for BNCT have already been pointed out earlier by Whittemore and have been verified in practice by the thermal neutron treatment work done at the Musashi 100 kW reactor. The advantages include a wide core face area and a wide spatial angle covered by the thermal-epithermal column system, large flux-per-Watt feature and inherent safety of the TRIGA fuel. Because of its wider applicability and less stringent requirements for clinical operation conditions, an epithermal neutron beam has been selected as the design goal. The epithermal flux should be sufficient for glioblastoma patient treatment: 109 epithermal neutrons/cm2/s with low enough fast neutron (-13Gy/epithermal n/cm2) and gamma contamination

  10. OPAL REACTOR: Calculation/Experiment comparison of Neutron Flux Mapping in Flux Coolant Channels

    Energy Technology Data Exchange (ETDEWEB)

    Barbot, L.; Domergue, C.; Villard, J. F.; Destouches, C. [CEA, Paris (France); Braoudakis, G.; Wassink, D.; Sinclair, B.; Osborn, J. C.; Huayou, Wu [ANSTO, Syeney (Australia)

    2013-07-01

    The measurement and calculation of the neutron flux mapping of the OPAL research reactor are presented. Following an investigation of fuel coolant channels using sub-miniature fission chambers to measure thermal neutron flux profiles, neutronic calculations were performed. Comparison between calculation and measurement shows very good agreement.

  11. Neutron spectra in two beam ports of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a 6Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

  12. Neutron spectra in two beam ports of the TRIGA Mark III reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R.; Hernandez D, V. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Aguilar, F.; Paredes, L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Rivera M, T., E-mail: fermineutron@yahoo.com [IPN, Centro de Investigacion en Ciencia Aplicada y Tecnologia Avanzada, Unidad Legaria, Av. Legaria 694, 11500 Mexico D. F. (Mexico)

    2013-10-15

    The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a {sup 6}Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

  13. New flux detectors for CANDU 6 reactors

    International Nuclear Information System (INIS)

    CANDU reactors utilize large numbers of in-core self-powered detectors for control and protection. In the original design, the detectors (coaxial cables) were wound on carrier tubes and immersed in the heavy water moderator. Failures occurred due to corrosion and other factors, and replacement was very costly because the assemblies were not designed with maintenance in mind. A new design was conceived based on straight detectors, of larger diameter, in a sealed package of individual 'well' tubes. This protected the detectors from hostile environments and enabled individual failed sensors to be replaced by inserting spares in vacant neighbouring tubes. The new design was made retrofittable to older CANDU reactors. Provision was made for on-line scanning of the core with a miniature fission chamber. The modified detectors were tested in a lengthy development program and found to exhibit superior performance to that of the original detectors. Most of the CANDU reactors have now adopted the new design. In the case of the Gentilly-2 and Point Lepreau reactors, advantage was taken of the opportunity to redesign the detector layout (using better codes and the increased flexibility in positioning detectors) to achieve better coverage of abnormal events, leading to higher trip setpoints and wider operating margins

  14. High frequency beam protection

    International Nuclear Information System (INIS)

    This report describes the design and construction of a high-frequency beam protecting device. This apparatus controls a number of functions in a modulator. Furthermore it drives the phase shifter and the attenuator. (author). 6 figs

  15. Split core experiments; Part I. Axial neutron flux distribution measurements in the reactor core with a central horizontal reflector

    International Nuclear Information System (INIS)

    A series of critical experiments were performed on the RB reactor in order to determine the thermal neutron flux increase in the central horizontal reflector formed by a split reactor core. The objectives of these experiments were to study the possibilities of improving the thermal neutron flux characteristics of the neutron beam in the horizontal beam tube of the RA research reactor. The construction of RA reactor enables to split the core in two, to form a central horizontal reflector in front of the beam tube. This is achieved by replacing 2% enriched uranium slugs in the fuel channel by dummy aluminium slugs. The purpose of the first series of experiments was to study the gain in thermal neutron component inside the horizontal reflector and the loss of reactivity as a function of the lattice pitch and central reflector thickness

  16. Computer code development for the analysis of reactivity transients using neutron transport theory and coupled thermofluid-dynamics for high-flux research reactors. Final report

    International Nuclear Information System (INIS)

    A new neutron transport code for time-dependent analyses of nuclear systems has been developed. The code system is based on the well-known discrete ordinates code DORT, which solves the steady-state neutron transport equation for an arbitrary number of energy groups and the most common regular 2D-geometries. For the implementation of time-dependence a fully implicit first-order time-integration scheme was employed to minimise errors due to temporal discretisation. This requires various modifications to the transport equation, the extensive use of sophisticated acceleration mechanisms and strongly tightened convergence criteria. To perform coupled analyses, an interface to the GRS (Gesellschaft fuer Reaktorsicherheit) system code ATHLET was developed. From the nodal power densities ATHLET calculates thermal-hydraulic system parameters, which are in turn used to generate appropriate nuclear cross sections. The code system has been applied to analyse the reactor dynamics of the research reactor FRM-II. After extensive steady-state analyses, several design basis accidents have been simulated: Loss of offsite power, loss of secondary heat sink and unintended withdrawl of control rod. Results from applying neutron transport theory were compared to diffusion theory and point kinetics calculations. Additionally, two hypothetical transients with large reactivity insertions of 3$ and 12% were considered to show the capability of the code system to cope with inhomogeneous reactivity insertions and strong flux distortions. (orig.)

  17. Experiments on critical heat flux for CAREM reactor

    International Nuclear Information System (INIS)

    The prediction of critical heat flux (CHF) in rod bundles of light water reactors is basically performed with the aid of empirical correlations derived from experimental data. Many CHF correlations have been proposed and are widely used in the analysis of the thermal margin during normal operation, transient, and accident conditions. Correlations found in the open literature are not sufficiently verified for the thermal-hydraulic conditions that appear in the CAREM core under normal operation: high pressure, low flow, and low qualities. To compensate this deficiency, an experimental investigation on CHF in such thermal-hydraulic conditions is being carried out. The experiments have been performed in the Institute of Physics and Power Engineering of Russian Federation. A short description of facilities, details of the experimental program and some trends in the preliminary results obtained are presented in this work. (author)

  18. Experiments on Critical Heat Flux for CAREM -25 Reactor

    International Nuclear Information System (INIS)

    The prediction of critical heat flux (CHF) in rod bundles of light water reactors is basically performed with the aid of empirical correlations derived from experimental data.Many CHF correlations have been proposed and are widely used in the analysis of the thermal margin during normal operation, transient, and accident conditions.Correlations found in the open literature are not sufficiently verified for the thermal hydraulic conditions that appear in the CAREM core under normal operation: high pressure, low flow, and low qualities.To compensate this deficiency, an experimental investigation on CHF in such thermal-hydraulic conditions was carried out.The experiments have been performed in the Institute of Physics and Power Engineering of Russian Federation.A short description of facilities, details of the experimental program and some preliminary results obtained are presented in this work

  19. Final beam transport in the reactor chamber

    International Nuclear Information System (INIS)

    The beam transport in heavy ion fusion (HIF) accelerators is discussed. The qualitative features of transport effects are presented. The basic transport effects associated with HIF beam are space charge effects, atomic physics effects, zero-order plasma effects, and plasma instabilities. In the case of HIF, very high intensity of HIF beam is required, and its own electric repulsion does not keep the beam converging. The number of beams required for supplying the demand power at a target can be estimated. The beam charge deposited on a target pellet produces electrostatic potential, and the electrostatic repulsion prevents the beam to reach on the target. The upper limit of the gas pressure is determined by small angle Coulomb scattering. Since unneutralized beam has the pinching force, the electrostatic kink mode (wiggle mode) should be considered in the pressure region where beam neutralization does not occur. Two-stream instability, filamentation instability and self-pinched transport are considered. As a conclusion of this paper, the new first choice for HIF transport is to use ballistic transport in moderate vacuum. (Kato, T.)

  20. Beam halo in high-intensity beams

    International Nuclear Information System (INIS)

    In space-charge dominated beams the nonlinear space-charge forces produce a filamentation pattern, which in projection to the 2-D phase spaces results in a 2-component beam consisting of an inner core and a diffuse outer halo. The beam-halo is of concern for a next generation of cw, high-power proton linacs that could be applied to intense neutron generators for nuclear materials processing. The author describes what has been learned about beam halo and the evolution of space-charge dominated beams using numerical simulations of initial laminar beams in uniform linear focusing channels. Initial results are presented from a study of beam entropy for an intense space-charge dominated beam

  1. Non-planar four-mirror optical cavity for high intensity gamma ray flux production by pulsed laser beam Compton scattering off GeV-electrons

    CERN Document Server

    Bonis, J; Cizeron, R; Cohen, M; Cormier, E; Cornebise, P; Delerue, N; Flaminio, R; Jehanno, D; Labaye, F; Lacroix, M; Marie, R; Mercier, B; Peinaud, Y; Pinard, L; Prevost, C; Soskov, V; Variola, A; Zomer, F

    2011-01-01

    As part of the R&D toward the production of high flux of polarised Gamma-rays we have designed and built a non-planar four-mirror optical cavity with a high finesse and operated it at a particle accelerator. We report on the main challenges of such cavity, such as the design of a suitable laser based on fiber technology, the mechanical difficulties of having a high tunability and a high mechanical stability in an accelerator environment and the active stabilization of such cavity by implementing a double feedback loop in a FPGA.

  2. Monitoring Akkuyu Nuclear Reactor Using Anti-Neutrino Flux Measurement

    CERN Document Server

    Ozturk, Sertac; Ozcan, V Erkcan; Unel, Gokhan

    2016-01-01

    We present a simulation based study for monitoring Akkuyu Nuclear Power Plant's activity using anti-neutrino flux originating from the reactor core. A water Cherenkov detector has been designed and optimization studies have been performed using Geant4 simulation toolkit. A first study for the design of a monitoring detector facility for Akkuyu Nuclear Power Plant has been discussed in this paper.

  3. Which reactor antineutrino flux may be responsible for the anomaly?

    CERN Document Server

    Giunti, Carlo

    2016-01-01

    We investigate which among the reactor antineutrino fluxes from the decays of the fission products of $^{235}\\text{U}$, $^{238}\\text{U}$, $^{239}\\text{Pu}$, and $^{241}\\text{Pu}$ may be responsible for the reactor antineutrino anomaly. We find that it is the $^{235}\\text{U}$ flux, which contributes to the rates of all reactor neutrino experiments. From the fit of the data we obtain the precise determination $ \\sigma_{^{235}\\text{U}} = ( 6.34 \\pm 0.10 ) \\times 10^{-43} \\, \\text{cm}^2 / \\text{fission} $ of the $^{235}\\text{U}$ cross section per fission, which is more precise than the calculated value and differs from it by $2.0\\sigma$.

  4. High-flux, extended-pulse accelerators: Final report

    International Nuclear Information System (INIS)

    The purpose of the program was to investigate physical phenomena associated with high flux ion beam generation and to develop technology for intense ion beam accelerators with pulselengths in the ms range. At the time the work was initiated, the chief area of application for the technology was ion implantation and materials modification

  5. Neutrino Flux Predictions for the NuMI Beam

    CERN Document Server

    Aliaga, L; Golan, T; Altinok, O; Bellantoni, L; Bercellie, A; Betancourt, M; Bravar, A; Budd, H; Carneiro, M F; Diaz, G A; Endress, E; Felix, J; Fields, L; Fine, R; Gago, A M; Galindo, R; Gallagher, H; Gran, R; Harris, D A; Higuera, A; Hurtado, K; Kiveni, M; Kleykamp, J; Le, T; Maher, E; Mann, W A; Marshall, C M; Caicedo, D A Martinez; McFarland, K S; McGivern, C L; McGowan, A M; Messerly, B; Miller, J; Mislivec, A; Morfin, J G; Mousseau, J; Naples, D; Nelson, J K; Norrick, A; Nuruzzaman,; Paolone, V; Park, J; Patrick, C E; Perdue, G N; Ransome, R D; Ray, H; Ren, L; Rimal, D; Rodrigues, P A; Ruterbories, D; Schellman, H; Salinas, C J Solano; Falero, S Sanchez; Tice, B G; Valencia, E; Walton, T; Wolcott, J; Wospakrik, M; Zhang, D

    2016-01-01

    Knowledge of the neutrino flux produced by the Neutrinos at the Main Injector (NuMI) beamline is essential to the neutrino oscillation and neutrino interaction measurements of the MINERvA, MINOS+, NOvA and MicroBooNE experiments at Fermi National Accelerator Laboratory. We have produced a flux prediction which uses all available and relevant hadron production data, incorporating measurements of particle production off of thin targets as well as measurements of particle yields from a spare NuMI target exposed to a 120 GeV proton beam. The result is the most precise flux prediction achieved for a neutrino beam in the one to tens of GeV energy region. We have also compared the prediction to in situ measurements of the neutrino flux and find good agreement.

  6. RHIC: The World's First High-Energy, Polarized-Proton Collider (423rd Brookhaven Lecture)

    International Nuclear Information System (INIS)

    The Relativistic Heavy Ion Collider (RHIC) at BNL has been colliding polarized proton at a beam energy of 100 billion electron volts (GeV) since 2001. In addition to reporting upon the progress of RHIC polarized-proton program, this talk will focus upon the mechanisms that cause the beam to depolarize and the strategies developed to overcome this. As the world first polarized-proton collider, RHIC is designed to collide polarized protons up to an energy of 250 GeV, thereby providing an unique opportunity to measure the contribution made by the gluon to a proton's spin and to study the spin structure of proton. Unlike other high-energy proton colliders, however, the challenge for RHIC is to overcome the mechanisms that cause partial or total loss of beam polarization, which is due to the interaction of the spin vector with the magnetic fields. In RHIC, two Siberian snakes have been used to avoid these spin depolarizing resonances, which are driven by vertical closed-orbit distortion and vertical betatron oscillations. As a result, polarized-proton beams have been accelerated to 100 GeV without polarization loss, although depolarization has been observed during acceleration from 100 GeV to 205 GeV.

  7. A High Flux Source of Cold Rubidium

    CERN Document Server

    Slowe, C; Hau, L V; Slowe, Christopher; Vernac, Laurent; Hau, Lene Vestergaard

    2004-01-01

    We report the production of a continuous, slow, and cold beam of 87-Rb atoms with an unprecedented flux of 3.2 x 10^12 atoms/s and a temperature of a few milliKelvin. Hot atoms are emitted from a Rb candlestick atomic beam source and transversely cooled and collimated by a 20 cm long atomic collimator section, augmenting overall beam flux by a factor of 50. The atomic beam is then decelerated and longitudinally cooled by Zeeman slowing.

  8. Status report of the program on neutron beam utilization at the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    The thermal reactor is an intense source not only of thermal neutron, but also intermediate as well as fast neutrons. Using the filtered neutron beam technique at steady state atomic reactor allows receiving the neutrons in the intermediate energy region with the most available intense flux at present. In the near time at the Dalat reactor the filtered neutron beam technique has been applied. Utilization of the filtered neutron beams in basic and applied researches has been a important activity of the Dalat Nuclear Research Institute (DNRI). This report presents some relevant characteristics of the filtered neutron beams and their utilization in nuclear data measurements, neutron capture gamma ray spectroscopy, neutron radiography, neutron dose calibration and other applications. (author). 3 refs, 2 figs

  9. The Brookhaven Radiation Effects Facility

    International Nuclear Information System (INIS)

    The Neutral Particle Beam (NPB) Radiation Effects Facility (REF), funded by the Strategic Defense Initiative Office (SDIO) through the Defense Nuclear Agency (DNA) and the Air Force Weapons Laboratory (AFWL), has been constructed at Brookhaven National Laboratory (BNL). Operation started in October 1986. The facility is capable of delivering pulsed H-, H/sup o/, and H+ beams of 100 to 200 MeV energy up to 30 mA peak current. Pulses can be adjusted from 5 μs to 500 μs length at a repetition rate of 5 pps. The beam spot on target is adjustable from 3 to 100 cm diameter (2 σ) resulting in a maximum dose of about 10 MRads (Si) per pulse (small beam spot). Experimental use of the REF is being primarily supported by the SDI lethality (LTH-4) program. The program has addressed ionization effects in electronics, both dose rate and total dose dependence, radiation-sensitive components, and dE/dx effects in energetic materials including propellants and high explosives (HE). This paper describes the facility, its capabilities and potential, and the experiments that have been carried out to date or are being planned. 2 refs., 10 figs

  10. High temperature gas reactor

    International Nuclear Information System (INIS)

    The present invention provides a reflector block structure of a high temperature gas reactor in which graphite blocks are not failed even a containing cylinder loaded to a fuel exchanger collides against to secured reflectors upon loading and withdrawing fuel constitutional elements. Namely, a protection plate made of a metal material such as stainless steel is covered on the secured reflector blocks disposed to the upper most step among secured graphite reflector blocks constituting the reactor core. In addition, positioning guide grooves are formed on the protection plate for guiding the containing cylinder loaded to the fuel exchanger to the column of the reactor core constitutional elements. With such a constitution, even if the containing cylinder of fuel exchanger is hoisted down and collided against the inner circumferential edge of the secured reflector blocks due to deviation of the position and the direction upon exchange of fuels, the reflector blocks are not failed since the above-mentioned portion is covered with the metal protection plate. In addition, the positioning guide grooves lead the fuel exchanger to a predetermined column correctly. (I.S.)

  11. PODESY program for flux mapping of CNA II reactor:

    International Nuclear Information System (INIS)

    The PODESY program, developed by KWU, calculates the spatial flux distribution of CNA II reactor through a three-dimensional expansion of 90 incore detector measurements. The calculation is made in three steps: a) short-term calculation which considers the control rod positions and it has to be done each time the flux mapping is calculated; b) medium-term calculation which includes local burn-up dependent calculation made by diffusion methods in macro-cell configurations (seven channels in hexagonal distribution), and c) long-term calculation, or macroscopic flux determination, that is a fitting and expansion of measured fluxes, previously corrected by local effects, using the eigen functions of the modified diffusion equation. The paper outlines development of step (c) of the calculation. The incore detectors have been located in the central zone of the core. In order to obtain low errors in the expansion procedure it is necessary to include additional points, whose flux values are assumed to be equivalent to detector measurements. These flux values are calculated with detector measurements and a spatial flux distribution calculated by a PUMA code. This PUMA calculation employs a smooth burn-up distribution (local burn-up variations are considered in step (b) of the whole calculation) representing the state of core evolution at the calculation time. The core evolution referred to ends when the equilibrium core condition is reached. Additionally, a calculation method to be employed in the plant in case of incore detector failures, is proposed. (Author)

  12. Future management of hazardous wastes generated at Brookhaven National Laboratory, Upton, New York. Environmental assessment

    International Nuclear Information System (INIS)

    This document assesses the potential environmental impacts of a variety of alternatives which could provide a Resource Conservation and Recovery Act (RCRA) permitted waste packaging and storage facility that would handle all hazardous, radioactive, and mixed wastes generated at Brookhaven National Laboratory (BNL) and would operate in full compliance with all federal, state, and local laws and regulations. Location of the existing Hazardous Waste Management Facility (HWMF) with respect to ground water and the site boundary, technical and capacity limitations, inadequate utilities, and required remediation of the area make the existing facility environmentally unacceptable for long term continued use. This Environmental Assessment (EA) describes the need for action by the Department of Energy (DOE). It evaluates the alternatives for fulfilling that need, including the alternative preferred by DOE, a no-action alternative, and other reasonable alternatives. The EA provides a general description of BNL and the existing environment at the current HWMF and alternative locations considered for a new Waste Management Facility (WMF). Finally, the EA describes the potential environmental impacts of the alternatives considered. The preferred alternative, also identified as Alternative D, would be to construct and operate a new WMF on land formerly occupied by barracks during Camp Upton operations, in an area north of Building 830 and the High Flux Beam Reactor/Alternating Gradient Synchrotron (AGS) recharge basins, east of North Railroad Street, and south of East Fifth Avenue. The purpose of this new facility would be to move all storage and transfer activities inside buildings and on paved and curbed areas, consolidate facilities to improve operations management, and provide improved protection of the environment

  13. Thermal flux flattering and increase of reactor output

    International Nuclear Information System (INIS)

    It is worthwhile, when building power reactors, to have excess reactivity in order to increase rating by fitting closely together the heat sources and the cooling possibilities. The power per unit volume of a graphite reactor can then be increased, given the power of the most heavily loaded channel. The solutions adopted for G.1, G.2, and E.D.F.1 are described here, and also the improvements based on the actual neutron flux flattening, the introduction of several zones for the coolant, the variation of uranium rod and coolant channel diameters according to their location, and finally the change in lattice pitch. The perturbation of neutron flux due to variation of mean absorption in the lattice is also discussed. (author)

  14. Charge exchange recombination spectroscopy measurements in the extreme ultraviolet region of central carbon concentrations during high power neutral beam heating in TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    The carbon concentration in the central region of TFTR discharges with high power neutral beam heating has been measured by charge-extracted recombination spectroscopy (CXRS) of the C+5 n = 3--4 transition in the extreme ultraviolet region. The carbon concentrations were deduced from absolute measurements of the line brightness using a calculation of the beam attenuation and the appropriate cascade-corrected line excitation rates. As a result of the high ion temperatures in most of the discharges, the contribution of beam halo neutrals to the line brightness was significant and therefore had to be included in the modeling of the data. Carbon concentrations have been measured in discharges with Ip = 1.0-1.6 MA and beam power in the range of 2.6-30 MW, including a number of supershots. The results are in good agreement with carbon concentrations deduced from the visible bremsstrahlung Zeff and metallic impurity concentrations measured by x-ray pulse-height analysis, demonstrating the reliability of the atomic rates used in the beam attenuation and line excitation calculations. Carbon is the dominant impurity species in these discharges; the oxygen concentration measured via CXRS in a high beam power case was 0.0006 of ne, compard to 0.04 for carbon. Trends with Ip and beam power in the carbon concentration and the inferred deuteron concentration are presented. The carbon concentration is independent of Ip and decreases from 0.13 at 2.6 MW beam power to 0.04 at 30 MW, while the deuteron concentration increases from 0.25 to 0.75 over the same range of beam power. These changes are primarily the result of beam particle fueling, as the carbon density did not vary significantly with beam power. The time evolutions of the carbon and deuteron concentrations during two high power beam pulses, one which exhibited a carbon bloom and one which did not, are compared. 30 refs., 12 figs., 2 tabs

  15. Modelling collimator of radial beam port Kartini reactor for boron neutron capture therapy

    International Nuclear Information System (INIS)

    One of the cancer therapy methods is BNCT (Boron Neutron Capture Therapy). BNCT utilizes neutron nature by 10B deposited on cancer cells. The superiority of BNCT compared to the radiation therapy is the high level of selectivity since its level is within cell. This study was carried out on collimator modelling in radial beam port of reactor Kartini for BNCT. The modelling was conducted by simulation using software of Monte Carlo N-Particle version 5 (MCNP 5). MCNP5 is a package of the programs for both simulating and calculating the problem of particle transport by following the life cycle of a neutron since its birth from fission reaction, transport on materials, until eventually lost due to the absorption reaction or out from the system. The collimator modelling used materials which varied in size in order to generate the value of each of the parameters in accordance with the recommendation of the IAEA, the epithermal neutron flux (ϕepi) > 1.0 x 109n.cm-2s-1, the ratio between the neutron dose rate fast and epithermal neutron flux (Df/ϕepi) < 2.0 x 10-13 Gy.cm2.n-1, the ratio of gamma dose rate and epithermal neutron flux (Dγ/ϕepi) < 2.0 X10-13 Gy.cm2.n-1, the ratio between the thermal and epithermal neutron flux (ϕTh/ϕepi)< 0.05 and the ratio between the current and flux of the epithermal neutron (J/ϕepi) > 0.7. Based on the results of the optimization of the modeling, the materials and sizes of the collimator construction obtained were 0.75 cm Ni as collimator wall, 22 cm Al as a moderator and 4.5 cm Bi as a gamma shield. The outputs of the radiation beam generated from collimator modeling of the radial beam port were ϕepi = 5.25 x 106 n.cm-2.s-1, Df/ϕepi = 1.17 x 10-13Gy.cm2.n-1, Dγ/ϕepi = 1.70 x 10-12 Gy.cm2.n-1, ϕTh/ϕepi = 1.51 and J/ϕepi = 0.731. Based on this study, the result of the beam radiation coming out of the radial beam port dis not fully meet the criteria recommended by IAEA so need to continue this study to get the criteria of IAEA

  16. High solids fermentation reactor

    Science.gov (United States)

    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.

  17. Brookhaven Highlights, January 1982-March 1983

    Energy Technology Data Exchange (ETDEWEB)

    Kuper, J.B.H.; Rustad, M.C. (eds.)

    1983-01-01

    Research at Brookhaven National Laboratory is summarized. Major headings are high energy physics, physics and chemistry, life sciences, applied energy science, support activities and administration. (GHT)

  18. Modeling space–time evolution of flux in a traveling wave reactor

    International Nuclear Information System (INIS)

    Highlights: • Monte-Carlo MCNPX was used to analyze flux profile in a traveling wave reactor. • Results show steady propagation of flux (2 cm/year) over life of the reactor. • High discharge burn-up of 394 GWd/MTU was observed for the prototype compact model. - Abstract: Simulations have been carried out using Monte Carlo code MCNPX to evaluate the space and time evolution of flux in a prototype traveling wave reactor under constant thermal power condition. A 3-D box-shaped model of the reactor is developed. The reactor core is divided into two primary regions: the smaller, enriched region with fissile material; and the larger non-enriched region with fertile material. This enrichment strategy is aimed to allow breed-and-burn in the core. The core, on the outside, is surrounded by shielding material of uniform thickness. To facilitate the study, these two primary regions in the core are further divided into thin slab-like regions referred to as cells. Results show propagation of flux profile from the enriched region to the non-enriched region at a near constant speed. Analyses of time evolution of local power density (power fraction) at specified locations in the core are presented. Space and time evolution of the overall core burn-up and localized burn-up are discussed

  19. Effects of high thermal and high fast fluences on the mechanical properties of Type 6061 aluminum on the HRBR

    International Nuclear Information System (INIS)

    The High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is an epithermal, externally moderated (by D2O) facility designed to produce neutron beams for research. Type 6061 T-6 aluminum was used for the beam tubes pressure vessel, fuel cladding, and most other components in the high flux area. The HFBR has operated since 1965. The epithermal, external moderation of the HFBR means that materials irradiated in different areas of the facility receive widely different flux spectra. Thus, specimens from a control rod drive follower tube (CRDF) have received 1.5 x 1022n/cm2(E > 0.1 MeV) and 3.2 x 1023n/cm2 thermal fluence, while those from a vertical thimble flow shroud received 1.9 x 1023n/cm2 (E > 0.1 MeV) and 1.0 x 1023n/cm2 thermal. These numbers correspond to fast to thermal fluence ratios ranging from 0.05 to 1.9. Irradiations are occurring at approximately 333 K. The data indicate that the increase in tensile strength and decrease in ductility result primarily from the thermal fluence, that is, transmutation of aluminum to silicon

  20. A wide range in-core neutron monitoring system for high powered TRIGA reactors

    International Nuclear Information System (INIS)

    High power movable core TRIGA reactors present unique problems of determining power levels from a neutron flux measurement because of (1) difficulty of locating detectors; (2) water thermal effects and (3) effect of experimental facilities. A solution, along with experimental results, will be described that uses a beam tube to effectively make in-core flux measurements with an out-of-core detector. The application of this new type of detector assembly to wide range linear and log power measurement will also be discussed. (author)

  1. Neutron Beam Characterization for Neutron Radiography Facility at the Thai Research Reactor TRR-1/M1

    International Nuclear Information System (INIS)

    The aim of this research is to characterize the present status of neutron beam coming out from the reactor core of Thai Research Reactor TRR-1/M1 through neutron radiography facility. In this study, the neutron beam profiles at different positions along the beam exit were recorded using digital imaging devices. In addition, thin foil activation technique, with and without cadmium cover, was employed to determine thermal neutron flux and Cd ratio. An acrylic step wedge was exposed to neutron at different time. In parallel to image construction, neutron detection was carried out using a BF3 gas-filled detector. Then, the image intensities at particular thicknesses were normalized by neutron counts from the BF3 detector to determine relative neutron intensity. The obtained information of neutron beam characterization will be useful not only for monitoring the present status of neutron radiography facility but also for determining the optimum exposure conditions for particular samples in the future.

  2. Application of reactor-pumped lasers to power beaming

    Science.gov (United States)

    Repetti, T. E.

    1991-10-01

    Power beaming is the concept of centralized power generation and distribution to remote users via energy beams such as microwaves or laser beams. The power beaming community is presently performing technical evaluations of available lasers as part of the design process for developing terrestrial and space-based power beaming systems. This report describes the suitability of employing a nuclear reactor-pumped laser in a power beaming system. Although there are several technical issues to be resolved, the power beaming community currently believes that the AlGaAs solid-state laser is the primary candidate for power beaming because that laser meets the many design criteria for such a system and integrates well with the GaAs photodiode receiver array. After reviewing the history and physics of reactor-pumped lasers, the advantages of these lasers for power beaming are discussed, along with several technical issues which are currently facing reactor-pumped laser research. The overriding conclusion is that reactor-pumped laser technology is not presently developed to the point of being technically or economically competitive with more mature solid-state technologies for application to power beaming.

  3. Extraterrestrial high energy neutrino fluxes

    International Nuclear Information System (INIS)

    With the aid of using the most recent cosmic ray spectra up to 2x1020 eV, production spectra of high-energy neutrinos from cosmic ray interactions with interstellar gas and extragalactic interactions of ultrahigh-energy cosmic rays with 3K universal background photons are presented and discussed. Estimates of the fluxes from cosmic diffuse sources and the nearby quasar 3C273 are made using the generic relationship between secondary neutrinos and gammas and using recent gamma ray satellite data. These gamma ray data provide important upper limits on cosmological neutrinos. Quantitative estimates of the observability of high-energy neutrinos from the inner galaxy and 3C273 above atmospheric background for a DUMAND-type detector are discussed in the context of the Weinberg-Salam model with sin2 theta/sub ω/ = 0.2 and including the atmospheric background from the decay of charmed mesons. Constraints on cosmological high-energy neutrino production models are also discussed. It appears that important high-energy neutrino astronomy may be possible with DUMAND, but very long observing times are required

  4. Extraterrestrial high energy neutrino fluxes

    Science.gov (United States)

    Stecker, F. W.

    1979-01-01

    Using the most recent cosmic ray spectra up to 2x10 to the 20th power eV, production spectra of high energy neutrinos from cosmic ray interactions with interstellar gas and extragalactic interactions of ultrahigh energy cosmic rays with 3K universal background photons are presented and discussed. Estimates of the fluxes from cosmic diffuse sources and the nearby quasar 3C273 are made using the generic relationship between secondary neutrinos and gammas and using recent gamma ray satellite data. These gamma ray data provide important upper limits on cosmological neutrinos. Quantitative estimates of the observability of high energy neutrinos from the inner galaxy and 3C273 above atmospheric background for a DUMAND type detector are discussed in the context of the Weinberg-Salam model with sq sin theta omega = 0.2 and including the atmospheric background from the decay of charmed mesons. Constraints on cosmological high energy neutrino production models are also discussed. It appears that important high energy neutrino astronomy may be possible with DUMAND, but very long observing times are required.

  5. Extraterrestrial high energy neutrino fluxes

    Energy Technology Data Exchange (ETDEWEB)

    Stecker, F.W.

    1979-06-01

    With the aid of using the most recent cosmic ray spectra up to 2x10/sup 20/ eV, production spectra of high-energy neutrinos from cosmic ray interactions with interstellar gas and extragalactic interactions of ultrahigh-energy cosmic rays with 3K universal background photons are presented and discussed. Estimates of the fluxes from cosmic diffuse sources and the nearby quasar 3C273 are made using the generic relationship between secondary neutrinos and gammas and using recent gamma ray satellite data. These gamma ray data provide important upper limits on cosmological neutrinos. Quantitative estimates of the observability of high-energy neutrinos from the inner galaxy and 3C273 above atmospheric background for a DUMAND-type detector are discussed in the context of the Weinberg-Salam model with sin/sup 2/ theta/sub ..omega../ = 0.2 and including the atmospheric background from the decay of charmed mesons. Constraints on cosmological high-energy neutrino production models are also discussed. It appears that important high-energy neutrino astronomy may be possible with DUMAND, but very long observing times are required.

  6. Advanced high temperature heat flux sensors

    Science.gov (United States)

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

    1983-01-01

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

  7. Study of neutron beam silhouette at tangential-through-tube of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Pakistan Research Reactor-1 (PARR-1) provides facilities to conduct experiments of vital importance using thermal neutron beams derived from the reactor core. One tangential-through-tube and several neutron beam tubes are available around the reactor for researchers. At the tangential-through-tube of PARR-1, experimental facilities for Prompt Gamma Neutron Activities Analysis (PGNAA) have been indigenously established. While designing the collimator it was imperative to ensure a proper collimation of thermal neutron beam on the target being exposed. It was, therefore, required to observe the neutron beam silhouette at various sections of the tangential-through-tube. In a series of experiments, CR-39 track detectors were exposed with neutrons at various sections of tangential-through-tube for about an hour while thermal neutron flux was measured in the range (1.8 x 10 /sup 7/ to 3.2 x 10 /sup 8/ neutrons cm/sup -2/.s/sup -1/. Optimum etching conditions were experimentally obtained to provide the best neutron beam profiles. The icon of the beam silhouette on the detectors can easily be observed with the naked eyes. However, innovative attempts have been made to reproduce the neutron silhouette onto paper by scanning these detectors using Laser jet scanner-4. This paper displays several scanned photographs of thermal neutron beam silhouette. In one of the neutron beam silhouettes, the neutron flux cut-off was successfully recognized. The neutron beam size was determined as -3.0 cm in diameter. This type of neutron beam silhouette study is not readily possible by other techniques. (author)

  8. Radiation protection commissioning of neutron beam instruments at the OPAL research reactor

    International Nuclear Information System (INIS)

    The neutron beam facilities at the 20 MW OPAL Research Reactor were commissioned in 2007 and 2008. The initial suite of eight neutron beam instruments on two thermal neutron guides, two cold neutron guides and one thermal beam port located at the reactor face, together with their associated shielding were progressively installed and commissioned according to their individual project plans. Radiation surveys were systematically conducted as reactor power was raised in a step-wise manner to 20 MW in order to validate instrument shielding design and performance. The performance of each neutron guide was assessed by neutron energy spectrum and flux measurements. The activation of beam line components, decay times assessments and access procedures for Bragg Institute beam instrument scientists were established. The multiple configurations for each instrument and the influence of operating more than one instrument or beamline simultaneously were also tested. Areas of interest were the shielding around the secondary shutters, guide shield and bunker shield interfaces and monochromator doors. The shielding performance, safety interlock checks, improvements, radiation exposures and related radiation protection challenges are discussed. This paper discusses the health physics experience of commissioning the OPAL Research Reactor neutron beam facilities and describes health physics results, actions taken and lessons learned during commissioning. (author)

  9. Radiation protection commissioning of neutron beam instruments at the OPAL Research Reactor

    International Nuclear Information System (INIS)

    The neutron beam facilities at the 20 MW OPAL Research Reactor were commissioned in 2007 and 2008. The initial suite of eight neutron beam instruments on two thermal neutron guides, two cold neutron guides and one thermal beam port located at the reactor face, together with their associated shielding were progressively installed and commissioned according to their individual project plans. Radiation surveys were systematically conducted as reactor power was raised in a step-wise manner to 20MW in order to validate instrument shielding design and performance. The performance of each neutron guide was assessed by neutron energy spectrum and flux measurements. The activation of beam line components, decay times assessments and access procedures for Bragg Institute beam instrument scientists were established. The multiple configurations for each instrument and the influence of operating more than one instrument or beamline simultaneously were also tested. Areas of interest were the shielding around the secondary shutters, guide shield and bunker shield interfaces and monochromator doors. The shielding performance, safety interlock checks, improvements, radiation exposures and related radiation protection challenges are discussed. This paper discusses the health physics experience of commissioning the OPAL Research Reactor neutron beam facilities and describes health physics results, actions taken and lessons learned during commissioning. (author)

  10. Design and Fabrication of a Highly Integrated Silicon Detector for the STAR Experiment at Brookhaven National Laboratory

    CERN Document Server

    Buck, Benjamin; Bessuille, Jason; Cepeda, Mario; Johnson, Thomas; Kelsey, James; van Nieuwenhuizen, Gerrit; Visser, Gerard

    2014-01-01

    We present the design of a detector used as a particle tracking device in the STAR experiment at the RHIC collider of Brookhaven National Laboratories. The "stave," 24 of which make up the completed detector, is a highly mechanically integrated design comprised of 6 custom silicon sensors mounted on a Kapton substrate. 4608 wire bonds connect these sensors to 36 analog front-end chips which are mounted on the same substrate. Power and signal connectivity from the hybrid to the front-end chips is provided by wire bonds. The entire circuit is mounted on a carbon fiber base co-cured to the Kapton substrate. We present the unique design challenges for this detector and some novel techniques for overcoming them.

  11. Thai Research Reactor (TRR-1/M1) Neutron Beam Measurements

    International Nuclear Information System (INIS)

    Full text: Neutron beam tube of neutron radiography facility at Thai Research Reactor (TRR-1/M1) Thailand Institute of Nuclear Technology (public organization) is a divergent beam. The rectangular open-end of the beam tube is 16 cm x 17 cm while the inner-end is closed to the reactor core. The neutron beam size was measured using 20 cm x 40 cm neutron imaging plate. The measurement at the position 100 cm from the end of the collimator has shown that the beam size was 18.2 cm x 19.0 cm. Gamma ray in neutron the beam was also measured by the identical position using industrial X ray film. The area of gamma ray was 27.8 cm x 31.1 cm with the highest intensity found to be along the neutron beam circumference

  12. BROOKHAVEN: Booster boost

    International Nuclear Information System (INIS)

    After three months of intensive dedicated machine studies, Brookhaven's new Booster accelerated 5 x 1013 protons over four cycles, about 85% of the design intensity. This was made possible by careful matching of Linac beam into the Booster and by extensive resonance stop band corrections implemented during Booster acceleration. The best single cycle injection into the AGS Alternating Gradient Synchrotron was 1.14 x 1013 protons from the Booster. 1.05 x 1013 protons were kept in the AGS, a 92% combined efficiency of extraction, transfer, and injection. The maximum injected 1994 shutdown period, enabling the 1994 physics run to make use of the full Booster intensity and go for the stated AGS objective of 4x1013 protons per pulse

  13. Critical heat flux prediction for the annular core research reactor

    International Nuclear Information System (INIS)

    This paper reports on best estimate predictions of Critical Heat Flux Ratio (CHFR) obtained to support the upgrade of the Annular Core Research Reactor (ACRR) at Sandia National Laboratories for 2 to 4 MWt. The CHF productions are based on the University of New Mexico's (UNM)-CHF correlations in conjunction with the Global Conditions Hypothesis (GCH). Results indicate that for the range of inlet water temperature of 293 to 333 K, CHFR predictions range from 3.9 to 2.1, which is more than sufficient to support the proposed ACRR upgrade

  14. Gamma-ray fluxes in Oklo natural reactors

    CERN Document Server

    Gould, C R; Sonzogni, A A; 10.1103/PhysRevC.86.054602

    2012-01-01

    Uncertainty in the operating temperatures of Oklo reactor zones impacts the precision of bounds derived for time variation of the fine structure constant $\\alpha$. Improved $^{176}$Lu/$^{175}$Lu thermometry has been discussed but its usefulness may be complicated by photo excitation of the isomeric state $^{176m}$Lu by $^{176}$Lu($\\gamma,\\gamma^\\prime $) fluorescence. We calculate prompt, delayed and equilibrium $\\gamma$-ray fluxes due to fission of $^{235}$U in pulsed mode operation of Oklo zone RZ10. We use Monte Carlo modeling to calculate the prompt flux. We use improved data libraries to estimate delayed and equilibrium spectra and fluxes. We find $\\gamma$-ray fluxes as a function of energy and derive values for the coefficients $\\lambda_{\\gamma,\\gamma^\\prime}$ that describe burn-up of $^{176}$Lu through the isomeric $^{176m}$Lu state. The contribution of the ($\\gamma,\\gamma^\\prime $) channel to the $^{176}$Lu/$^{175}$Lu isotopic ratio is negligible in comparison to the neutron burn-up channels. Lutetium...

  15. Heat transfer phenomena in gas protected particle beam fusion reactor cavities

    International Nuclear Information System (INIS)

    The behavior of the fireball produced in particle beam fusion reactor cavities as the cavity gas near the target absorbs the X-rays and ionic debris emanating from the microexplosion is examined. Thermal response of the first wall to the radiative heat flux from the gas is examined parametrically. Criteria for the suitability of different cavity fill gases based on their ability to protect the first wall from excessive surface heating and ablation are discussed. 9 refs

  16. Calculation of neutron fluxes in biological shield of the TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    The complete calculation of neutron fluxes in biological shield and verification with experimental results is presented. Calculated results are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in irradiation channel 4 in TRIGA Mark II reactor. These experimental results were used as a benchmark. Homogeneous type of problem (without inserted irradiation channel) and problem with asymmetry (inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. Deviation from material data set up as original parameters is also considered (first of all presence of water in concrete and density of concrete) for type of concrete in biological shield and for selected type of concrete in irradiation channel. BUGLE-96 (47 neutron energy groups) library is used. Excellent agreement between calculated and experimental results for reaction rate is received.(author)

  17. Magnetoinduction converter for measuring the charged particle flux in beams

    International Nuclear Information System (INIS)

    The arrangement of a contactless magnetoinduction converter (MIC) designed for measuring the charged particle flux in beams is described. The converter is made of a coil placed onto a toroidal ferromagnetic core, 120x60x12 mm in size. To eliminate the effect of the external magnetic field the MIC is placed into a compound permalloy- copper labyrinth-type screen, In the aperiodic operating mode the MIC measuring channel contains a preamplifier, an amplifier, a strobing circuit, an integrator with a converter, a delay circuit, a time relay, a pulsed-to-direct voltage converter, and a digital voltmeter. For experimental measuring of sensitivity of the MIC measuring system a calibration loop, consisting of an accurate- amplitude generator, a delay circuit and a time relay, is used. The given contactless magnetoinduction converter is a part of the electron flux standard for 5-50 MeV beams. The normal conditions of reproduction of the ''electron/s'' unit are the following: the 293+-1 K temperature, 101.3+40 kPa pressure, 60+-15% relative humidity, 220B+-10% supply voltage and 50+-0.5 Hz frequency. The dynamic range of MIC application is 1012-1015 electron/s. The total systematic error of reproduction of the electron flux unit for the MIC is 1.7%

  18. Evaluation of critical heat flux performances for design strategy of new research reactor nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Bang, In Cheol; Lee, Kwi Lim; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2006-02-15

    The present project investigated stable burnout heat flux correlations applicable to research reactor operation conditions of low pressure, low temperature and high flow rate. In addition, in series of thermal limits important to safety of the reactor, ONB and OFI correlations also were investigated. There are some world CHF databases for tube-inside flow. In order to design a research reactor, DNB is final design limit factor and so the collection of the data or correlation are very important. The optimal core cooling capability can be done by considering neutronics, economical efficiency, materials limit together through engineering judgement based on DNB correlations. The project collected the materials and correlations applicable to research reactor conditions. The correlations give a fundamental base for analyzing thermal limit factors and will be used helpfully in review of regulatory body and designer for safety evaluation.

  19. Muon flux measurement with silicon detectors in the CERN neutrino beams

    International Nuclear Information System (INIS)

    The present work mainly describes the 'Neutrino Flux Monitoring' system (NFM), which has been built for the 400-GeV Super Proton Synchrotron (SPS) neutrino beams. A treatment is given of some general subjects related to the utilization of silicon detectors and the properties of high-energy muons. Energy loss of minimal-ionizing particles, which has to be distinguished from energy deposition in the detector, is considered. Secondary radiation, also called 'spray', consisting of 'delta rays' and other cascade products, is shown to play an important role in the muon flux measurement inside a shield, especially for muons of high energy (> 100 GeV). Radiation induced damage in the detectors, which determines the long term performance, is discussed. The relation between the detector response and the real muon flux is determined. The use of NFM system for on-line beam monitoring is described. (Auth.)

  20. Study of the RP-10 reactor neutron beam applied to the neutron radiography

    International Nuclear Information System (INIS)

    We have studied the RP-10 reactor radial neutron beam No. 3, which is used for neutron radiographies, by comparing radiograph's with and without the inner duct, and neutron flux determination with in flakes along the external duct, being the presence of photons creating signals at comparable levels of neutron effects, which reduce the quality of the analysis, values around 106 and 104 n/cm2s for thermal and epithermal flux were obtained respectively. It is recommended evaluate the design of the internal duct which presents strong photon emission. (authors).

  1. Neutron beam facilities at the Replacement Research Reactor, ANSTO

    International Nuclear Information System (INIS)

    The exciting development for Australia is the construction of a modern state-of-the-art 20-MW Replacement Research Reactor which is currently under construction to replace the aging reactor (HIFAR) at ANSTO in 2006. To cater for advanced scientific applications, the replacement reactor will provide not only thermal neutron beams but also a modern cold-neutron source moderated by liquid deuterium at approximately -250 deg C, complete with provision for installation of a hot-neutron source at a later stage. The latest 'supermirror' guides will be used to transport the neutrons to the Reactor Hall and its adjoining Neutron Guide Hall where a suite of neutron beam instruments will be installed. These new facilities will expand and enhance ANSTO's capabilities and performance in neutron beam science compared with what is possible with the existing HIFAR facilities, and will make ANSTO/Australia competitive with the best neutron facilities in the world. Eight 'leading-edge' neutron beam instruments are planned for the Replacement Research Reactor when it goes critical in 2006, followed by more instruments by 2010 and beyond. Up to 18 neutron beam instruments can be accommodated at the Replacement Research Reactor, however, it has the capacity for further expansion, including potential for a second Neutron Guide Hall. The first batch of eight instruments has been carefully selected in conjunction with a user group representing various scientific interests in Australia. A team of scientists, engineers, drafting officers and technicians has been assembled to carry out the Neutron Beam Instrument Project to successful completion. Today, most of the planned instruments have conceptual designs and are now being engineered in detail prior to construction and procurement. A suite of ancillary equipment will also be provided to enable scientific experiments at different temperatures, pressures and magnetic fields. This paper describes the Neutron Beam Instrument Project and gives

  2. Microdosimetry of high LET therapeutic beams

    International Nuclear Information System (INIS)

    Experimental microdosimetry of high LET therapeutic beams were presented. The cyclotron produced fast neutron beams at IMS, TAMVEC and NRL, a reactor fast neutron at YAYOI, a proctor beam at Harvard and a pion beam at TRIUMF are included. Measurements were performed with a conventional tissue equivalent spherical proportional counter with a logarithmic amplifier which made the recording and analysis quite simple. All the energy deposition spectra were analysed in the conventional manner and anti y F, anti y D as well as anti y D* were calculated. The spectra and their mean lineal energies showed wide variations, depending on the particle type, energy, position in phantom. Fractional contribution of elemental particles ( electron, muon, pion, proton, alpha and so on) to the total dose were analysed. For fast neutron beams, the y spectra stayed almost constant at any depth along the central axis in the phantom. The y spectra of proton beam changed slightly along the depth. On the other side, the y spectra of pion beam change drastically in the phantom between plateau and dose peak region. A novel technique of time-of-flight microdosimetry was employed, which made it possible to separate the fractional contribution of contaminant electrons and muons out of pions. Finally, a map of the radiation quality for all the beams is presented and its significances are discussed. (author)

  3. Neutron spectrum measurements at a radial beam port of the NUR research reactor using a Bonner spheres spectrometer.

    Science.gov (United States)

    Mazrou, H; Nedjar, A; Seguini, T

    2016-08-01

    This paper describes the measurement campaign held around the neutron radiography (NR) facility of the Algerian 1MW NUR research reactor. The main objective of this work is to characterize accurately the neutron beam provided at one of the radial channels of the NUR research reactor taking benefit of the acquired CRNA Bonner spheres spectrometer (BSS). The specific objective was to improve the image quality of the NR facility. The spectrometric system in use is based on a central spherical (3)He thermal neutron proportional counter combined with high density polyethylene spheres of different diameters ranging from 3 to 12in. This counting system has good gamma ray discrimination and is able to cover an energy range from thermal to 20MeV. The measurements were performed at the sample distance of 0.6m from the beam port and at a height of 1.2m from the facility floor. During the BSS measurements, the reactor was operating at low power (100W) to avoid large dead times, pulse pileup and high level radiation exposures, in particular, during spheres handling. Thereafter, the neutron spectrum at the sample position was unfolded by means of GRAVEL and MAXED computer codes. The thermal, epithermal and fast neutron fluxes, the total neutron flux, the mean energy and the Cadmium ratio (RCd) were provided. A sensitivity analysis was performed taking into account various defaults spectra and ultimately a different response functions in the unfolding procedure. Overall, from the obtained results it reveals, unexpectedly, that the measured neutron spectrum at the sample position of the neutron radiography of the NUR reactor is being harder with a predominance of fast neutrons (>100keV) by about 60%. Finally, those results were compared to previous and more recent measurements obtained by activation foils detectors. The agreement was fairly good highlighting thereby the consistency of our findings. PMID:27203706

  4. Flux flow and flux dynamics in high-Tc superconductors

    Science.gov (United States)

    Bennett, L. H.; Turchinskaya, M.; Swartzendruber, L. J.; Roitburd, A.; Lundy, D.; Ritter, J.; Kaiser, D. L.

    1991-01-01

    Because high temperature superconductors, including BYCO and BSSCO, are type 2 superconductors with relatively low H(sub c 1) values and high H(sub c 2) values, they will be in a critical state for many of their applications. In the critical state, with the applied field between H(sub c 1) and H(sub c 2), flux lines have penetrated the material and can form a flux lattice and can be pinned by structural defects, chemical inhomogeneities, and impurities. A detailed knowledge of how flux penetrates the material and its behavior under the influence of applied fields and current flow, and the effect of material processing on these properties, is required in order to apply, and to improve the properties of these superconductors. When the applied field is changed rapidly, the time dependence of flux change can be divided into three regions, an initial region which occurs very rapidly, a second region in which the magnetization has a 1n(t) behavior, and a saturation region at very long times. A critical field is defined for depinning, H(sub c,p) as that field at which the hysteresis loop changes from irreversible to reversible. As a function of temperature, it is found that H(sub c,p) is well described by a power law with an exponent between 1.5 and 2.5. The behavior of H(sub c,p) for various materials and its relationship to flux flow and flux dynamics are discussed.

  5. Flux flow and flux dynamics in high-Tc superconductors

    International Nuclear Information System (INIS)

    Because high temperature superconductors, including BYCO and BSSCO, are type 2 superconductors with relatively low H(sub c 1) values and high H(sub c 2) values, they will be in a critical state for many of their applications. In the critical state, with the applied field between H(sub c 1) and H(sub c 2), flux lines have penetrated the material and can form a flux lattice and can be pinned by structural defects, chemical inhomogeneities, and impurities. A detailed knowledge of how flux penetrates the material and its behavior under the influence of applied fields and current flow, and the effect of material processing on these properties, is required in order to apply, and to improve the properties of these superconductors. When the applied field is changed rapidly, the time dependence of flux change can be divided into three regions, an initial region which occurs very rapidly, a second region in which the magnetization has a 1n(t) behavior, and a saturation region at very long times. A critical field is defined for depinning, H(sub c,p) as that field at which the hysteresis loop changes from irreversible to reversible. As a function of temperature, it is found that H(sub c,p) is well described by a power law with an exponent between 1.5 and 2.5. The behavior of H(sub c,p) for various materials and its relationship to flux flow and flux dynamics are discussed

  6. Axial flux distribution in a lattice position in the NRX reactor

    International Nuclear Information System (INIS)

    The axial thermal flux distribution in a lattice position in the NRX reactor has been measured at a number of moderator levels. The results have been fitted to sine functions and values are given for the positions of the flux maxima and the extrapolated flux lengths. Results of measurements of the axial fast flux distribution are also given. (author)

  7. High quality flux control system for electron gun evaporation

    International Nuclear Information System (INIS)

    This paper reports on a high quality flux control system for electron gun evaporation developed and tested for the MBE growth of high temperature superconductors. The system can be applied to any electron gun without altering the electron gun itself. Essential elements of the system are a high bandwidth mass spectrometer, control electronics and a high voltage modulator to sweep the electron beam over the melt at high frequencies. the sweep amplitude of the electron beam is used to control the evaporation flux at high frequencies. The feedback loop of the system has a bandwidth of over 100 Hz, which makes it possible to grow superlattices and layered structures in a fast and precisely controlled manner

  8. Materials research with neutron beams from a research reactor

    International Nuclear Information System (INIS)

    Because of the unique ways that neutrons interact with matter, neutron beams from a research reactor can reveal knowledge about materials that cannot be obtained as easily with other scientific methods. Neutron beams are suitable for imaging methods (radiography or tomography), for scattering methods (diffraction, spectroscopy, and reflectometry) and for other possibilities. Neutron-beam methods are applied by students and researchers from academia, industry and government to support their materials research programs in several disciplines: physics, chemistry, materials science and life science. The arising knowledge about materials has been applied to advance technologies that appear in everyday life: transportation, communication, energy, environment and health. This paper illustrates the broad spectrum of materials research with neutron beams, by presenting examples from the Canadian Neutron Beam Centre at the NRU research reactor in Chalk River. (author)

  9. Materials research with neutron beams from a research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Root, J.; Banks, D. [Canadian Neutron Beam Centre, Chalk River Laboratories, Chalk River, Ontario (Canada)

    2015-03-15

    Because of the unique ways that neutrons interact with matter, neutron beams from a research reactor can reveal knowledge about materials that cannot be obtained as easily with other scientific methods. Neutron beams are suitable for imaging methods (radiography or tomography), for scattering methods (diffraction, spectroscopy, and reflectometry) and for other possibilities. Neutron-beam methods are applied by students and researchers from academia, industry and government to support their materials research programs in several disciplines: physics, chemistry, materials science and life science. The arising knowledge about materials has been applied to advance technologies that appear in everyday life: transportation, communication, energy, environment and health. This paper illustrates the broad spectrum of materials research with neutron beams, by presenting examples from the Canadian Neutron Beam Centre at the NRU research reactor in Chalk River. (author)

  10. Survey of research reactors

    International Nuclear Information System (INIS)

    A survey of reasearch reactors based on the IAEA Nuclear Research Reactor Data Base (RRDB) was done. This database includes information on 273 operating research reactors ranging in power from zero to several hundred MW. From these 273 operating research reactors 205 reactors have a power level below 5 MW, the remaining 68 reactors range from 5 MW up to several 100 MW thermal power. The major reactor types with common design are: Siemens Unterrichtsreaktors, 1.2 Argonaut reactors, Slowpoke reactors, the miniature neutron source reactors, TRIGA reactors, material testing reactors and high flux reactors. Technical data such as: power, fuel material, fuel type, enrichment, maximum neutron flux density and experimental facilities for each reactor type as well as a description of their utilization in physics and chemistry, medicine and biology, academic research and teaching, training purposes (students and physicists, operating personnel), industrial application (neutron radiography, silicon neutron transmutation doping facilities) are provided. The geographically distribution of these reactors is also shown. As conclusions the author discussed the advantages (low capital cost, low operating cost, low burn up, simple to operate, safe, less restrictive containment and sitting requirements, versatility) and disadvantages (lower sensitivity for NAA, limited radioisotope production, limited use of neutron beams, limited access to the core, licensing) of low power research reactors. 24 figs., refs. 15, Tab. 1 (nevyjel)

  11. High flux transmutation of fission products and actinides

    International Nuclear Information System (INIS)

    Long-lived fission products and minor actinides accumulated in spent nuclear fuel of power reactors comprise the major part of high level radwaste. Their incineration is important from the point of view of radwaste management. Transmutation of these nuclides by means of neutron irradiation can be performed either in conventional nuclear reactors, or in specialized transmutation reactors, or in ADS facilities with subcritical reactor and neutron source with application of proton accelerator. Different types of transmutation nuclear facilities can be used in order to insure optimal incineration conditions for radwaste. The choice of facility type for optimal transmutation should be based on the fundamental data in the physics of nuclide transformations. Transmutation of minor actinides leads to the increase of radiotoxicity during irradiation. It takes significant time compared to the lifetime of reactor facility to achieve equilibrium without effective transmutation. High flux nuclear facilities allow to minimize these draw-backs of conventional facilities with both thermal and fast neutron spectrum. They provide fast approach to equilibrium and low level of equilibrium mass and radiotoxicity of transmuted actinides. High flux facilities are advantageous also for transmutation of long-lived fission products as they provide short incineration time

  12. ITER relevant high heat flux testing on plasma facing surfaces

    International Nuclear Information System (INIS)

    The current ITER design employs beryllium, carbon fiber reinforced composite and tungsten as plasma facing materials. Since these materials are exposed to high heat fluxes during the operation, it is essential to perform high heat flux for R and D of ITER components. Static heat loads corresponding to cycling loads during normal operation, are estimated to be up to 20 MW/m2 in the divertor targets and around 0.5 MW/m2 at the first wall in ITER. For the static high heat flux testing, tests in electron beam facilities, particle beam facilities, IR heater and in-pile tests have been performed. Another type, more critical heat loads, which have high power densities and short durations, corresponding to transient events, i.e. plasma disruption, vertical displacement events (VDEs) and edge localized modes (ELMs) deliver considerable heat flux onto the plasma facing materials. For this purpose, tests in electron beam (short pulses), plasma gun and high power laser facilities have been carried out. The present work summarizes the features of these facilities and recent experimental results as well as the current selection of ITER plasma facing components. (author)

  13. High Pressure Boiling Water Reactor

    International Nuclear Information System (INIS)

    Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits are; 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel is smooth and is without penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340 oC (15 MPa) steam instead of ∼285 oC (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Moisture separators and steam dryers are inside the reactor vessel and additional separators and dryers can be installed inside or outside the containment -Well proved simple dry containment or wet containment can be used. (author)

  14. Present status of neutron beam facilities at the research reactor, HANARO, and its future prospect

    International Nuclear Information System (INIS)

    Korea has been operating its new research reactor, HANARO, since its first criticality in 1995. It is an open-tank-in-pool type reactor using LEU fuel with thermal neutron flux of 2 x 1014 nominally at the nose in the D2O reflector having 7 horizontal beam ports and a provision of vertical hole for cold neutron source installation. KAERI has pursued an extensive instrument development program since 1992 by the support of the nuclear long-term development program of the government and there are now 4 working instruments. A high resolution powder diffractometer and a neutron radiography facility has been operational since late 1997 and 1996, respectively. A four-circle diffractometer has been fully working since mid 1999 and a small angle neutron spectrometer is just under commissioning phase. With the development of linear position sensitive detector with delay-line readout electronics, we have developed a residual stress instrument as an optional machine to the HRPD for last two years. Around early 1998 informal users program started with friendly users and it became a formal users support program by the ministry of science and technology. Short description for peer group formation and users activities is given. (author)

  15. Present status of neutron beam facilities at the research reactor, HANARO, and its future prospect

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang-Hee; Kang, Young-Hwan; Kuk, Il-Hiun [Korea Atomic Energy Research Institute, Taejon (Korea)

    2001-03-01

    Korea has been operating its new research reactor, HANARO, since its first criticality in 1995. It is an open-tank-in-pool type reactor using LEU fuel with thermal neutron flux of 2 x 10{sup 14} nominally at the nose in the D{sub 2}O reflector having 7 horizontal beam ports and a provision of vertical hole for cold neutron source installation. KAERI has pursued an extensive instrument development program since 1992 by the support of the nuclear long-term development program of the government and there are now 4 working instruments. A high resolution powder diffractometer and a neutron radiography facility has been operational since late 1997 and 1996, respectively. A four-circle diffractometer has been fully working since mid 1999 and a small angle neutron spectrometer is just under commissioning phase. With the development of linear position sensitive detector with delay-line readout electronics, we have developed a residual stress instrument as an optional machine to the HRPD for last two years. Around early 1998 informal users program started with friendly users and it became a formal users support program by the ministry of science and technology. Short description for peer group formation and users activities is given. (author)

  16. Analysis of calculated neutron flux response at detectors of G.A. Siwabessy multipurpose reactor (RSG-GAS Reactor)

    International Nuclear Information System (INIS)

    Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect

  17. Improved performance over time of integration in momentum flux estimation using Postset Beam Steering technique

    Science.gov (United States)

    Anandan, V. K.; Kumar, Shridhar; Sureshbabu, V. N.; Rao, T. Narayana; Rao, M. Purnachandra; Tsuda, Toshitaka

    2014-12-01

    Using time series of vertical data collected from middle and upper atmospheric (MU) radar, the adaptive Capon beamforming technique is used to synthesize beams in the desired pointing directions within the radar beamwidth. Beam synthesis has been performed at the tilt angle of 1.5° with different beam configurations (4 beams, 8 beams, 16 beams, 32 beams, 48 beams and 64 beams), which are equally separated azimuth plane. The Power Spectral Density (PSD) is obtained from the synthesized beam using the eigenvector (EV) sub-space based spectral estimation method. The first moment is derived from the EV produced spectrum using the adaptive moment estimation method. From the first order moments derived along equally spaced pointing directions, radial wind velocities are readily obtained in the corresponding directions. From radial velocity obtained in various pointing directions, momentum flux of short duration (<2 h) is estimated using the symmetric beam method (SBM) which requires 4 symmetric beams separated by 90° in the azimuth plane. A comparative study has been performed to study the optimum beam configuration, required for the flux estimation, which shows that 32 beam configuration is sufficient to estimate the flux with least error. Study with 32 beam configuration, gives 8 sets of symmetric beams (8×4). Using 8 set of beams, the flux is estimated for each set of beams and averaged. The averaged flux is further integrated over different lengths of time up to 14 h. This systematic method for the estimation of momentum flux reveals that spatial averaging of beams in azimuth and integration over different lengths of time have reduced the time of integration from 15 to 16 h for the conventional approach to 8-9 h for the new approach using Postset Beam Steering (PBS) technique. The flux estimated with spatial-averaging of beam using PBS technique has been compared with other standard methods.

  18. Neutron beam applications using low power research reactor Malaysia perspectives

    International Nuclear Information System (INIS)

    The TRIGA MARK II Research reactor at the Malaysian Institute for Nuclear Research (MINT) was commissioned in July 1982. Since then various works have been performed to utilise the neutrons produced from this steady state reactor. One area currently focussed on is the utilisation of neutron beam ports available at this 1MW reactor. Projects undertaken are the development and utilisation of the Neutron Radiography (myNR), Small Angle Neutron Scattering (mySANS) and Boron Neutron Capture Therapy (BNCT) - preliminary study. In order to implement active research programmes, a group comprised of researcher from research institutes and academic institutions, has formed: known as Malaysian Reactor Interest Group (MRIG). This paper describes the recent status the above neutron beam facilities and their application in industrial, health and material technology research and education. The related activities of MRIG are also highlighted. (author)

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

    Science.gov (United States)

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

    1994-10-01

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

  20. A Level 1+ Probabilistic Safety Assessment of the high flux Australian reactor. Vol. 2. Appendix C: System analysis models and results

    International Nuclear Information System (INIS)

    This section contains the results of the quantitative system/top event analysis. Section C. 1 gives the basic event coding scheme. Section C.2 shows the master frequency file (MFF), which contains the split fraction names, the top events they belong to, the mean values of the uncertainty distribution that is generated by the Monte Carlo quantification in the System Analysis module of RISKMAN, and a brief description of each split fraction. The MFF is organized by the systems modeled, and within each system, the top events associated with the system. Section C.3 contains the fault trees developed for the system/top event models and the RISKMAN reports for each of the system/top event models. The reports are organized under the following system headings: Compressed/Service Air Supply (AIR); Containment Isolation System (CIS); Heavy Water Cooling System (D20); Emergency Core Cooling System (ECCS; Electric Power System (EPS); Light Water Cooling system (H20); Helium Gas System (HE); Mains Water System (MW); Miscellaneous Top Events (MISC); Operator Actions (OPER) Reactor Protection System (RPS); Space Conditioner System (SCS); Condition/Status Switch (SWITCH); RCB Ventilation System (VENT); No. 1 Storage Block Cooling System (SB)

  1. A Level 1+ Probabilistic Safety Assessment of the high flux Australian reactor. Vol. 2. Appendix C: System analysis models and results

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    This section contains the results of the quantitative system/top event analysis. Section C. 1 gives the basic event coding scheme. Section C.2 shows the master frequency file (MFF), which contains the split fraction names, the top events they belong to, the mean values of the uncertainty distribution that is generated by the Monte Carlo quantification in the System Analysis module of RISKMAN, and a brief description of each split fraction. The MFF is organized by the systems modeled, and within each system, the top events associated with the system. Section C.3 contains the fault trees developed for the system/top event models and the RISKMAN reports for each of the system/top event models. The reports are organized under the following system headings: Compressed/Service Air Supply (AIR); Containment Isolation System (CIS); Heavy Water Cooling System (D20); Emergency Core Cooling System (ECCS); Electric Power System (EPS); Light Water Cooling system (H20); Helium Gas System (HE); Mains Water System (MW); Miscellaneous Top Events (MISC); Operator Actions (OPER) Reactor Protection System (RPS); Space Conditioner System (SCS); Condition/Status Switch (SWITCH); RCB Ventilation System (VENT); No. 1 Storage Block Cooling System (SB)

  2. Brookhaven National Laboratory ADS concepts (USA)

    International Nuclear Information System (INIS)

    Three types of accelerator driven system have been investigated at Brookhaven National Laboratory: the accelerator driven energy producer (ADEP) which is intended for energy production, incineration of minor actinides (MA) and long-lived fission products (LLFP) using a small power accelerator; the Phoenix concept aimed at transmuting large amounts of MA and LLFP using a high current linear accelerator; the accelerator driven particle fuel transmutor (ADPF) which also transmutes the MA and LLFP at high rate by using a high neutron flux. The paper presents these three concepts and other related problems such as fuel and coolant materials, targets, subcriticality and safety issues, the use of Thorium cycle, and non-proliferation issue

  3. The high temperature reactor

    International Nuclear Information System (INIS)

    The outstanding characteristics of the HTR could not save this well-proved sort from an eventful fate. They did ensure, however, that despite all hindrances and delays, extensive experiences with the HTR are available today which form a broad fundament for this type which is regarded as very important for the future. The most important starting point for the future might be the fact that now all industrial groups interested in the HTR have gathered for a joint deliberation of the situation and decision making. The guiding lines seem to be drawn already: there are broad fields of application on the heat market, but power generation also offers incentives, connected with reasons of fuel-saving. All things considered, today's world energy situation represents the biggest challenge and the most powerful impetus to this reactor which possesses the important basic lines of the innovation required in energy converting processes. (orig.)

  4. Hydrogen production from fusion reactors coupled with high temperature electrolysis

    International Nuclear Information System (INIS)

    The decreasing availability of fossil fuels emphasizes the need to develop systems which will produce synthetic fuel to substitute for and complement the natural supply. An important first step in the synthesis of liquid and gaseous fuels is the production of hydrogen. Thermonuclear fusion offers an inexhaustible source of energy for the production of hydrogen from water. Processes which may be considered for this purpose include electrolysis, thermochemical decomposition or thermochemical-electrochemical hybrid cycles. Preliminary studies at Brookhaven indicate that high temperature electrolysis has the highest potential efficiency for production of hydrogen from fusion. Depending on design electric generation efficiencies of approximately 40 to 60 percent and hydrogen production efficiencies of approximately 50 to 70 percent are projected for fusion reactors using high temperature blankets

  5. Estimation of impact pressure due to rupture in beam-tube for research reactor

    International Nuclear Information System (INIS)

    Neutrons have been used for studies in material sciences of physics, chemistry, metals and alloys, ceramics, polymers, and biological sciences. This application leads to build up research reactor all over the world. JRTR (Jordan Research and Training Reactor) which plans to build up in Jordan is multipurpose research reactor which is developed entirely with domestic technology to overseas. Thermal power is 5MW upgradable 10MW. JRTR have four horizontal beam tubes, 3 ST(Standard) and 1NR (Neutron radiography). The beam tube's cavities are filled with helium, purged regularly to prevent a build-up of radioactive gases and moisture. They are highly reliable because they have no moving parts. The beam tube embedded part is aligned with its corresponding beam tube in the reflector. Objective of this study is to describe water hammer phenomenon in beam tube and determine an impact pressure charged in end film of beam tube for accomplishing nuclear safety function of research reactor while beam tube is ruptured due to some accident such as earthquake. The water hammer was experimentally and analytically studied by Lai, Saruba, Ballanco, and Watters

  6. High βp bootstrap tokamak reactor

    International Nuclear Information System (INIS)

    Basic characteristics of a steady state tokamak fusion reactor is presented. The minimum required energy multiplication factor Q is found to be 20 to 30 for the feasibility of the fusion reactor. Such a high Q steady state tokamak operation is possible, within our present knowledge of the operational constraints and the current drive physics, when a large fraction of the plasma current is carried by the bootstrap current. Operation at high βp (≥2.0) and high qψ (=4-5) with relatively small εβp (3) and fusion output power (2.5 GW) and is consistent with the present knowledges of the plasma physics of the tokamak, namely the Troyon limit, the energy confinement scalings, the bootstrap current, the current drive efficiency (NB current drive with the total power of 70 MW and the beam energy of 1 MeV) with a favorable aspect on the formation of the cold and dense diverter plasma-condition. From the economical aspect of the tokamak fusion reactor, a more compact reactor is favorable. The use of the high field magnet with Bmax = 16T (for example Ti-doped Nb3Sn conductor) enables to reduce the total machine size to 50% of the above-described conventional design, namely Rp = 7m, Vp = 760m-3, PF = 2.8 GW. (author)

  7. UCN sources at external beams of thermal neutrons. An example of PIK reactor

    Science.gov (United States)

    Lychagin, E. V.; Mityukhlyaev, V. A.; Muzychka, A. Yu.; Nekhaev, G. V.; Nesvizhevsky, V. V.; Onegin, M. S.; Sharapov, E. I.; Strelkov, A. V.

    2016-07-01

    We consider ultracold neutron (UCN) sources based on a new method of UCN production in superfluid helium (4He). The PIK reactor is chosen as a perspective example of application of this idea, which consists of installing 4He UCN source in the beam of thermal or cold neutrons and surrounding the source with moderator-reflector, which plays the role of cold neutron (CN) source feeding the UCN source. CN flux in the source can be several times larger than the incident flux, due to multiple neutron reflections from the moderator-reflector. We show that such a source at the PIK reactor would provide an order of magnitude larger density and production rate than an analogous source at the ILL reactor. We estimate parameters of 4He source with solid methane (CH4) or/and liquid deuterium (D2) moderator-reflector. We show that such a source with CH4 moderator-reflector at the PIK reactor would provide the UCN density of ~1·105 cm-3, and the UCN production rate of ~2·107 s-1. These values are respectively 1000 and 20 times larger than those for the most intense UCN user source. The UCN density in a source with D2 moderator-reflector would reach the value of ~2·105 cm-3, and the UCN production rate would be equal ~8·107 s-1. Installation of such a source in a beam of CNs would slightly increase the density and production rate.

  8. Modernized IBR-2 reactor and first experiments at its neutron beams

    International Nuclear Information System (INIS)

    IBR-2 reactor is the main basic facility at JINR dedicated to condensed matter research. The IBR-2 operates as a fast pulsed reactor. Its main distinctive property, which makes it differ from other nuclear reactors, is the mechanical modulation of the reactivity by means of a movable reflector. Producing a record neutron flux of 1016 n/cm2/s in the pulse, the IBR-2 reactor is also an economical and relatively inexpensive facility. The IBR-2 reactor is mainly used for investigations in the fields of condensed matter physics (solids and liquids), biology, chemistry, earth and materials science. Operating experience has shown that it is a very effective neutron source; in most areas of application it compares well with the best neutron sources based on proton accelerators. At present, this experience is of special importance in connection with the increasing interest in long-pulsed neutron sources. IBR-2 operated successfully from 1984 until 2006. On December 18, 2006 reactor was shut down for modernization. Main directions of reactor modernization include: 1. A compact reactor core. 2. Lower speed of rotation of the main movable reflector, counter rotation of rotors, use of a nickel alloy as a reflector material. 3. Use fuel pellets configuration that will allow increasing the depth of fuel burn up to 9%. 4. New design of safety system which improves its parameters. 5. Creation of easily replaceable moderators, their optimization for each neutron beam. Development of the cryogenic moderators with palletized moderator material. Dismantling of the old reactor parts and installation of the new equipment was completed in 2010. Brief history of the work will be outlined in the report. In February 2011 loading of fresh fuel to the reactor core was completed and physical start-up has begun. After successful realization of this stage in June, the power start-up program was fulfilled resulting in increase of the mean reactor power to the design value of 2 MW (peak power of

  9. Investigating The Neutron Flux Distribution Of The Miniature Neutron Source Reactor MNSR Type

    International Nuclear Information System (INIS)

    Neutron flux distribution is the important characteristic of nuclear reactor. In this article, four energy group neutron flux distributions of the miniature neutron source reactor MNSR type versus radial and axial directions are investigated in case the control rod is fully withdrawn. In addition, the effect of control rod positions on the thermal neutron flux distribution is also studied. The group constants for all reactor components are generated by the WIMSD code, and the neutron flux distributions are calculated by the CITATION code. The results show that the control rod positions only affect in the planning area for distribution in the region around the control rod. (author)

  10. Muon flux measurement with silicon detectors in the CERN neutrino beams

    International Nuclear Information System (INIS)

    The neutrino beam installations at the CERN SPS accelerator are described, with emphasis on the beam monitoring systems. Especially the muon flux measurement system is considered in detail, and the calibration procedure and systematic aspects of the measurements are discussed. An introduction is given to the use of silicon semiconductor detectors and their related electronics. Other special chapters concern non-linear phenomena in the silicon detectors, radiation damage in silicon detectors, energy loss and energy deposition in silicon and a review of energy loss phenomena for high energy muons in matter. (orig.)

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

    Science.gov (United States)

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

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

  12. A high temperature reactor for ship propulsion

    International Nuclear Information System (INIS)

    The initial thermal hydraulic and physics design of a high temperature gas cooled reactor for ship propulsion is described. The choice of thermodynamic cycle and thermal power is made to suit the marine application. Several configurations of a Helium cooled, Graphite moderated reactor are then analysed using the WIMS and MONK codes from AEA Technology. Two geometries of fuel elements formed using micro spheres in prismatic blocks, and various arrangements of control rods and poison rods are examined. Reactivity calculations through life are made and a pattern of rod insertion to flatten the flux is proposed and analysed. Thermal hydraulic calculations are made to find maximum fuel temperature under high power with optimized flow distribution. Maximum temperature after loss of flow and temperatures in the reactor vessel are also computed. The temperatures are significantly below the known limits for the type of fuel proposed. It is concluded that the reactor can provide the required power and lifetime between refueling within likely space and weight constraints. (author)

  13. Absolute beam flux measurement at NDCX-I using gold-melting calorimetry technique

    International Nuclear Information System (INIS)

    We report on an alternative way to measure the absolute beam flux at the NDCX-I, LBNL linear accelerator. Up to date, the beam flux is determined from the analysis of the beam-induced optical emission from a ceramic scintilator (Al-Si). The new approach is based on calorimetric technique, where energy flux is deduced from the melting dynamics of a gold foil. We estimate an average 260 kW/cm2 beam flux over 5 (micro)s, which is consistent with values provided by the other methods. Described technique can be applied to various ion species and energies.

  14. Neutron flux determinations in the reactors G2 and G3 during operation

    International Nuclear Information System (INIS)

    After demonstrating the sensitivity of the distribution of power in a production reactor to a deformation caused by dissymmetries of reactivity in the reactor, the authors describe the method of neutron flux determination devised for the reactors G2 and G3 under working conditions; the detector used is a tungsten or nickel wire, the γ activity of which is measured with an ionisation chamber. Several flux determinations are given as examples to illustrate the sensitivity of the method. (author)

  15. Experimental estimation of the neutron flux density at the reconstructed Rossendorf research reactor

    International Nuclear Information System (INIS)

    The Rossendorf Research Reactor was reconstructed in the years 1986-1989. During start up of the reactor the neutron flux density was investigated in the reactor core and the outer irradiation channels by the help of activation probes and self-powered neutron detectors. The report includes the most important experimental results and a brief description of the measuring techniques. (orig.)

  16. Determination of neutronic fluxes in research nuclear reactor of Triga Mark I and WWRS types

    International Nuclear Information System (INIS)

    In this paper is presented the determination of the thermal, epithermal and fast neutron fluxes, using neutron activation analysis technique, for two research nuclear reactors of different design: the Triga Mark I reactor was designed by Gulf General Atomic Co in USA and the WWRS reactor was designed in the URSS, both in the 50's years. (Author)

  17. Spatial fluxes and energy distributions of reactor fast neutrons in two types of heat resistant concretes

    International Nuclear Information System (INIS)

    Measurements have been carried out to study the spatial fluxes and energy distributions of reactor fast neutrons transmitted through two types of heat resistant concretes, serpentine concrete and magnetic lemonite concrete. The physical, chemical and mechanical properties of these concretes were checked by well known techniques. In addition, the effect of heating at temperatures up to 500deg C on the crystaline water content was checked by the method of differential thermal analysis. Measurements were performed using a collimated beam of reactor neutrons emitted from a 10 MW research reactor. The neutron spectra transmitted through concrete barriers of different thickness were measured by a scintillation spectrometer with NE-213 liquid organic scintillator. Discrimination against undesired pulses due to gamma-rays was achieved by a method based on pulse shape discrimination technique. The operating principle of this technique is based on the comparison of two weighted time integrals of the detector signal. The measured pulse amplitude distribution was converted to neutron energy distribution by a computational code based on double differentiation technique. The spectrometer workability and the accuracy of the unfolding technique were checked by measuring the neutron spectra of neutrons from Pu-α-Be and 252Cf neutron sources. The obtained neutron spectra for the two concretes were used to derive the total cross sections for neutrons of different energies. (orig.)

  18. Numerical effects in the neutron flux calculations into WWER type reactors vessels by Monte Carlo Method

    International Nuclear Information System (INIS)

    The calculation of neutron fluxes and fluence into reactor pressure vessel is a regulatory requirement in the stages of the design, operation and plan lifetime extension. Being the reactor vessel a part of the primary circuit, its integrity should be preserved under all operation regimes. The reactor vessel is considered a unique and non-substitutable part of the NPP that undergoes degradation. The main source of the aging comes from the fast neutron damage induced in the steel crystalline lattice. In the case of the WWER-type reactors, the vessel fragilization has been identified as one of the main problems concerning the safety of NPPs. Due to the proximity of the core edge to the vessel inner surface; the vessel steel is exposed to high fast neutron fluence. The effect of this irradiation on the mechanical properties becomes more acute because of the impurities measured in the current Russian steel alloys. In the present paper, a PC version of the Monte Carlo 3-D HEXANN-EVALU system is used for the estimation of the WWER reactor pressure vessel irradiation. It was selected on the basis of its flexible options that on the other hand need to be quantified in connection with the desired magnitudes. The parameters that control the random walk of neutrons as well as the efficiency increasing options included in the code are studied in order to identify their impact in the final results for fluxes and fluence in the reactor pressure vessel. As a result an optimal set of parameters is suggested

  19. High flux expansion divertor studies in NSTX

    CERN Document Server

    Soukhanovskii, V A; Bell, R E; Gates, D A; Kaita, R; Kugel, H W; LeBlanc, B P; Maqueda, R; Menard, J E; Mueller, D; Paul, S F; Raman, R; Roquemore, A L

    2009-01-01

    High flux expansion divertor studies have been carried out in the National Spherical Torus Experiment using steady-state X-point height variations from 22 to 5-6 cm. Small-ELM H-mode confinement was maintained at all X-point heights. Divertor flux expansions from 6 to 26-28 were obtained, with associated reduction in X-point connection length from 5-6 m to 2 m. Peak divertor heat flux was reduced from 7-8 MW/m$^2$ to 1-2 MW/m$^2$. In low X-point configuration, outer strike point became nearly detached. Among factors affecting deposition of parallel heat flux in the divertor, the flux expansion factor appeared to be dominant

  20. Survey of the thermal and fast neutron flux distribution in the core of IPR-R1 reactor

    International Nuclear Information System (INIS)

    A methodology to obtain the neutron flux distribution inside the core of a reactor is presented, aiming to analyze specifications for increasing reactor power. The activation measurement technique with irradiation of steel eletrodes of 700 mm of lenght, put in acrylic rods was used. In the detection process and in the counting of activation product, a Ge (Li) detector with high resolution and a scanning mechanical system, constructed and projected in CDTN (Nuclear Technology Development Center) were used. (E.G.)

  1. Fuel cladding integrity analysis during beam trip transients for China lead-based demonstration reactor

    International Nuclear Information System (INIS)

    Highlights: • Beam trip effect on Accelerator Driven sub-critical System (ADS) is remained a critical issue on ADS reactor technology. • The CFD model of fuel pin of China Lead-based Demonstration Reactor (CLEAR-III) was established. • The thermal hydraulic behaviors of fuel pin during beam trip transient of CLEAR-III were studied. • The thermal stress variation of fuel cladding during beam trip transient of CLEAR-III was evaluated. • Results reveal that beam trip effect on fuel cladding is so small that can be neglected. - Abstract: Frequent beam trips as experienced in the existing high-power proton accelerators may cause thermal fatigue in Accelerator-Driven System (ADS) components, which may lead to degradation of their structural integrity and reduction of their lifetime. In this paper, we focus on the strength and integrity of fuel cladding during the beam trip transients of China Lead-based Demonstration Reactor (CLEAR-III). Typical frequent beam trips and fuel burn-up are addressed to investigate the acceptable beam trip frequency limitation. Correspondingly, the variation magnitude of temperature and thermal stress of fuel cladding are simulated by ANSYS code. Besides, the behavior of cladding material T91 under irradiation, creep and Lead Bismuth Eutectic (LBE) corrosion conditions has been discussed. It shows that beam trips have little influence on the cladding integrity and the acceptable beam trip frequency of the fuel cladding within 10 s of the beam trip time duration is more than 2.5 × 105 times per year, consequently the CLEAR-III’s fuel claddings are expected to have a good resistance to the thermal–mechanical effects induced by beam trips

  2. Results from post-mortem tests with materials from the old core-box of the High Flux Reactor (HFR) at Petten

    International Nuclear Information System (INIS)

    Results are reported from hardness measurements, tensile tests and fracture mechanics experiments (fatigue crack growth and fracture toughness) on 5154 aluminum specimens fabricated from remnants of the old HFR core box. The specimen material was exposed to a maximum thermal neutron fluence of 7.5 * 1026 n/m2(E 26n/m2, but with a thermal to fast neutron ratio of about 4, shows more radiation hardening : 67HR15N, 0.2 - yield strength 580 MPa and 1.5% total elongation. Fatigue crack growth rates range from 5 * 10-5mm/cycle to 10-3mm/cycle for ΔK ranging from 8 to 20 MPa√m. The most highly exposed (7.5 * 10 26n/m2) materials shows accelerated fatigue crack growth due to unstable crack extension at ΔK of about 15 MPa√m. The lowermost meaningful measure of plane strain fracture toughness is 18 MPa√m. Except for the fracture toughness, which is a factor of about 3 higher, the results show reasonable agreement with the expected mechanical properties estimated in the 'safe end-of-life' assessment of the old HFR vessel

  3. Dissolution of low burnup Fast Flux Test reactor fuel

    International Nuclear Information System (INIS)

    The first Fast-Flux Test Facility reactor fuel [mixed (U,Pu)O2 composition] has been used in dissolution tests for fuel reprocessing. The fuel tested here had a peak burnup of 0.22 at. %, with peak centerline temperatures of 19970C. Linear dissolution rates of 0.99 to 1.57 mm/h were determined for dissolver solution and fresh acid, respectively. Insoluble residues from dissolution at 950C ranged from 0.18 to 0.28% of the original fuel. From 2 to 37 wt % of the residue was recoverable plutonium. Dissolution at 290C yielded residues of 0.56 to 0.64% of the original fuel. The major elements present in the HF leached residue included Ru, Mo, and Rh. The recovered cladding from the 950C dissolution contained the equivalent of 198 mg of 239Pu per 100 g of hulls, while the cladding from the 290c experiments contained only 0.21 mg of 239Pu per 100 g of hulls. 9 references, 5 figures

  4. Brookhaven fastbus/unibus interface

    Energy Technology Data Exchange (ETDEWEB)

    Benenson, G.; Bauernfeind, J.; Larsen, R.C.; Leipuner, L.B.; Morse, W.M.; Adair, R.K.; Black, J.K.; Campbell, S.R.; Kasha, H.; Schmidt, M.P.

    1983-01-01

    A typical high energy physics experiment requires both a high speed data acquisition and processing system, for data collection and reduction; and a general purpose computer to handle further reduction, bookkeeping and mass storage. Broad differences in architecture, format or technology, will often exist between these two systems, and interface design can become a formidable task. The PDP-11 series minicomputer is widely used in physics research, and the Brookhaven FASTBUS is the only standard high speed data acquisition system which is fully implemented in a current high energy physics experiment. This paper will describe the design and operation of an interface between these two systems. The major issues are elucidated by a preliminary discussion on the basic principles of Bus Systems, and their application to Brookhaven FASTBUS and UNIBUS.

  5. Brookhaven fastbus/unibus interface

    International Nuclear Information System (INIS)

    A typical high energy physics experiment requires both a high speed data acquisition and processing system, for data collection and reduction; and a general purpose computer to handle further reduction, bookkeeping and mass storage. Broad differences in architecture, format or technology, will often exist between these two systems, and interface design can become a formidable task. The PDP-11 series minicomputer is widely used in physics research, and the Brookhaven FASTBUS is the only standard high speed data acquisition system which is fully implemented in a current high energy physics experiment. This paper will describe the design and operation of an interface between these two systems. The major issues are elucidated by a preliminary discussion on the basic principles of Bus Systems, and their application to Brookhaven FASTBUS and UNIBUS

  6. Barrier Cavities in the Brookhaven AGS

    International Nuclear Information System (INIS)

    In collaboration with KEK two barrier cavities, each generating 40 kV per turn have been installed in the Brookhaven AGS. Machine studies are described and their implications for high intensity operations are discussed

  7. 2012 review of French research reactors

    International Nuclear Information System (INIS)

    Proposed by the French Reactor Operators' Club (CER), the meeting and discussion forum for operators of French research reactors, this report first gives a brief presentation of these reactors and of their scope of application, and a summary of highlights in 2012 for each of them. Then, it proposes more detailed presentations and reviews of characteristics, activities, highlights, objectives and results for the different types of reactors: neutron beam reactors (Orphee, High flux reactor-Laue-Langevin Institute or HFR-ILL), technological irradiation reactors (Osiris and Phenix), training reactors (Isis and Azur), reactors for safety research purposes (Cabri and Phebus), reactors for neutronic studies (Caliban, Prospero, Eole, Minerve and Masurca), and new research reactors (the RES facility and the Jules Horowitz reactor or JHR)

  8. Estimation of reactivity effect of neutron beam tube in research reactor through two-dimensional transport calculation. Comparison of radial and tangential beam tubes

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Masatoshi

    1988-12-01

    The reactivity effects of neutron beam tubes in a research reactor were investigated with the two-dimensional transport code DORT. The core model for the calculation was a two-dimensional cylinder. The reactivity effects of one radial and two tangential beam tubes were estimated through the results of the two-dimensional calculation using the space-dependent weight function which is as defined a product of the macroscopic scattering cross section, the forward neutron flux, the adjoint neutron flux and the volume. The reactivity effect of the tangential beam tube is larger than that of the radial tube. An aluminum wall of a beam tube decreases the reactivity of the core due to the neutron absorption.

  9. Monte Carlo simulation of the thermal column and beam tube of the TRIGA Mark II research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R., E-mail: rustamzia@yahoo.com [Atominstitute (ATI), Vienna University of Technology (TU Wien), Stadion allee 2, A-1020 Vienna (Austria); Karimzadeh, S.; Stummer, T.; Boeck, H. [Atominstitute (ATI), Vienna University of Technology (TU Wien), Stadion allee 2, A-1020 Vienna (Austria)

    2011-08-15

    Highlights: > Neutronics parameters of the reactor shielding. > Biological shielding of the TRIGA reactor. > Thermal flux measurement in the thermal column and BT-A. > MCNP model validation. - Abstract: The Monet Carlo simulation of the TRIGA Mark II research reactor core has been performed employing the radiation transport computer code MCNP5. The model has been confirmed experimentally in the PhD research work at the Atominstitute (ATI) of the Vienna University of Technology. The MCNP model has been extended to complete biological shielding of the reactor including the thermal column, radiographic collimator and four beam tubes. This paper presents the MCNP simulated results in the thermal column and one of the beam tubes (beam tube A) of the reactor. To validate these theoretical results, thermal neutron flux density measurements using the gold foil activation method have been performed in the thermal column and beam tube A (BT-A). In the thermal column, the theoretical and experimental results are in fairly good agreement i.e. maximum thermal flux density in the centre decreases in radial direction. Further, it is also agreed that thermal flux densities in the lower part is greater than the upper part of the thermal column. In the BT-A experiment, the thermal flux density distribution is measured using gold foil. The experimental and theoretical diffusion lengths have been determined as 10.77 cm and 9.36 cm respectively with only 13% difference, reflecting good agreement between the experimental and simulated results. To save the computational cost and to incorporate the accurate and complete information of each individual Monte Carlo MC particle tracks, the surface source writing capability of MCNP has been utilized to the TRIGA shielding model. The variance reduction techniques have been applied to improve the statistics of the problem and to save computational efforts.

  10. Monte Carlo simulation of the thermal column and beam tube of the TRIGA Mark II research reactor

    International Nuclear Information System (INIS)

    Highlights: → Neutronics parameters of the reactor shielding. → Biological shielding of the TRIGA reactor. → Thermal flux measurement in the thermal column and BT-A. → MCNP model validation. - Abstract: The Monet Carlo simulation of the TRIGA Mark II research reactor core has been performed employing the radiation transport computer code MCNP5. The model has been confirmed experimentally in the PhD research work at the Atominstitute (ATI) of the Vienna University of Technology. The MCNP model has been extended to complete biological shielding of the reactor including the thermal column, radiographic collimator and four beam tubes. This paper presents the MCNP simulated results in the thermal column and one of the beam tubes (beam tube A) of the reactor. To validate these theoretical results, thermal neutron flux density measurements using the gold foil activation method have been performed in the thermal column and beam tube A (BT-A). In the thermal column, the theoretical and experimental results are in fairly good agreement i.e. maximum thermal flux density in the centre decreases in radial direction. Further, it is also agreed that thermal flux densities in the lower part is greater than the upper part of the thermal column. In the BT-A experiment, the thermal flux density distribution is measured using gold foil. The experimental and theoretical diffusion lengths have been determined as 10.77 cm and 9.36 cm respectively with only 13% difference, reflecting good agreement between the experimental and simulated results. To save the computational cost and to incorporate the accurate and complete information of each individual Monte Carlo MC particle tracks, the surface source writing capability of MCNP has been utilized to the TRIGA shielding model. The variance reduction techniques have been applied to improve the statistics of the problem and to save computational efforts.

  11. Plasma–Surface Interactions Under High Heat and Particle Fluxes

    Directory of Open Access Journals (Sweden)

    Gregory De Temmerman

    2013-01-01

    Full Text Available The plasma-surface interactions expected in the divertor of a future fusion reactor are characterized by extreme heat and particle fluxes interacting with the plasma-facing surfaces. Powerful linear plasma generators are used to reproduce the expected plasma conditions and allow plasma-surface interactions studies under those very harsh conditions. While the ion energies on the divertor surfaces of a fusion device are comparable to those used in various plasma-assited deposition and etching techniques, the ion (and energy fluxes are up to four orders of magnitude higher. This large upscale in particle flux maintains the surface under highly non-equilibrium conditions and bring new effects to light, some of which will be described in this paper.

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

    Science.gov (United States)

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

    2015-10-01

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

  13. Dosimetric characteristics of the thermal neutron beam facility for neutron capture therapy at Hanaro reactor

    International Nuclear Information System (INIS)

    The thermal neutron beam facility utilizing a typical tangential beam port for Neutron Capture Therapy was installed at the Hanaro, 30 MW multi-purpose research reactor. In order to determine the different dose components in phantoms irradiated with a mixed thermal neutron beam and gamma-ray for clinical applications, various techniques were applied including the use of activation foils, TLDs and ionization chambers. The water phantom was utilized in the measurement. The results of the measurement were compared with MCNP4B calculations. The thermal neutron fluxes were 1.02E9 and 6.07E8/cm2·s at 10 and 20 mm depth in water, respectively. The gamma-ray dose rate was 5.10 Gy/hr at 20 mm depth in water. The result of this study can be used as basic data for subsequent BNCT clinical application. (author)

  14. Steel slag carbonation in a flow-through reactor system: the role of fluid-flux.

    Science.gov (United States)

    Berryman, Eleanor J; Williams-Jones, Anthony E; Migdisov, Artashes A

    2015-01-01

    Steel production is currently the largest industrial source of atmospheric CO2. As annual steel production continues to grow, the need for effective methods of reducing its carbon footprint increases correspondingly. The carbonation of the calcium-bearing phases in steel slag generated during basic oxygen furnace (BOF) steel production, in particular its major constituent, larnite {Ca2SiO4}, which is a structural analogue of olivine {(MgFe)2SiO4}, the main mineral subjected to natural carbonation in peridotites, offers the potential to offset some of these emissions. However, the controls on the nature and efficiency of steel slag carbonation are yet to be completely understood. Experiments were conducted exposing steel slag grains to a CO2-H2O mixture in both batch and flow-through reactors to investigate the impact of temperature, fluid flux, and reaction gradient on the dissolution and carbonation of steel slag. The results of these experiments show that dissolution and carbonation of BOF steel slag are more efficient in a flow-through reactor than in the batch reactors used in most previous studies. Moreover, they show that fluid flux needs to be optimized in addition to grain size, pressure, and temperature, in order to maximize the efficiency of carbonation. Based on these results, a two-stage reactor consisting of a high and a low fluid-flux chamber is proposed for CO2 sequestration by steel slag carbonation, allowing dissolution of the slag and precipitation of calcium carbonate to occur within a single flow-through system. PMID:25597686

  15. Fast current pulse amplifier for neutron flux monitoring system of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    The neutron flux monitoring system (NFMS) for Prototype Fast Breeder Reactor (PFBR) measures the neutron power and the reactivity changes in the core in all the states such as shut down, fuel handling, reactor startup, intermediate and power ranges using high temperature cylindrical fission chambers, four section fission counter and high temperature boron coated counter. Fast Current Pulse Amplifier has been developed to use in NFMS of PFBR that amplifies single/four numbers of input current pulses independently, discriminates and electronically wire - OR them to give differential pulse output along with the Campbell output. The paper describes the design, development of integrated single/Quad channel fast current pulse amplifier based on in-house developed ASIC, Hybrid IC, in built test features, LV and HV supplies. (author)

  16. High energy beam manufacturing technologies

    International Nuclear Information System (INIS)

    Technological progress continues to enable us to utilize ever widening ranges of physical and chemical conditions for material processing. The increasing cost of energy, raw materials and environmental control make implementation of advanced technologies inevitable. One of the principal avenues in the development of material processing is the increase of the intensity, accuracy, flexibility and stability of energy flow to the processing site. The use of different forms of energy beams is an effective way to meet these sometimes incompatible requirements. The first important technological applications of high energy beams were welding and flame cutting. Subsequently a number of different kinds of beams have been used to solve different problems of part geometry control and improvement of surface characteristics. Properties and applications of different specific beams were subjects of a number of fundamental studies. It is important now to develop a generic theory of beam based manufacturing. The creation of a theory dealing with general principles of beam generation and beam-material interaction will enhance manufacturing science as well as practice. For example, such a theory will provide a format approach for selection and integration of different kinds of beams for a particular application. And obviously, this theory will enable us to integrate the knowledge bases of different manufacturing technologies. The War of the Worlds by H. G. Wells, as well as a number of more technical, although less exciting, publications demonstrate both the feasibility and effectiveness of the generic approach to the description of beam oriented technology. Without any attempt to compete with Wells, we still hope that this volume will contribute to the creation of the theory of beam oriented manufacturing

  17. High flux isotope reactor technical specifications

    International Nuclear Information System (INIS)

    Technical specifications are presented concerning safety limits and limiting safety system settings; limiting conditions for operation; surveillance requirements; design features; administrative controls; and accidents and anticipated transients

  18. Measurement of (n,xn) reaction cross-sections using prompt {gamma} spectroscopy at neutron beams with high instantaneous flux; Mesure de sections efficaces de reaction (n,xn) par spectroscopie {gamma} prompte aupres d'un faisceau a tres haut flux instantane

    Energy Technology Data Exchange (ETDEWEB)

    Lukic, S

    2004-10-15

    The work presented in this thesis is situated in the context of the GEDEON program of neutron cross-section measurements. This program is motivated by the perspectives recently opened by projects of nuclear waste treatment and energy production. There is an obvious lack of experimental data on (n,xn) reactions in the databases, especially in the case of very radioactive isotopes. An important technique to measure cross-sections of these reactions is the prompt {gamma}-ray spectroscopy at white pulsed neutron beams with very high instantaneous flux. In this work, inelastic scattering and (n,xn) reactions cross-section measurements were performed on a lead sample from threshold to 20 MeV by prompt {gamma}-ray spectroscopy at the white neutron beam generated by GELINA facility in Geel, Belgium. Digital methods were developed to treat HPGe CLOVER detector signals and separate {gamma}-rays induced by the fastest neutrons from those belonging to the flash. Partial cross-sections for the production of several transitions in natural lead were measured and analyzed using theoretical calculations in order to separate the contributions of different reactions leading to the same residual isotope. Total cross-sections of the reactions in question were estimated. The results were compared to the TALYSS code theoretical calculations, as well as to other experimental results. This experiment has served to validate the method and it opens the way to measure (n,xn) reactions cross-sections with high instantaneous neutron flux on actinides, particularly the U{sup 233}(n,2n) reaction which is important for the thorium cycle. (author)

  19. Filtered epithermal quasi-monoenergetic neutron beams at research reactor facilities

    International Nuclear Information System (INIS)

    Filtered neutron techniques were applied to produce quasi-monoenergetic neutron beams in the energy range of 1.5–133 keV at research reactors. A simulation study was performed to characterize the filter components and transmitted beam lines. The filtered beams were characterized in terms of the optimal thickness of the main and additive components. The filtered neutron beams had high purity and intensity, with low contamination from the accompanying thermal emission, fast neutrons and γ-rays. A computer code named “QMNB” was developed in the “MATLAB” programming language to perform the required calculations. - Highlights: • Quasi-monoenergetic neutron beams in energy range from (1.5–133) keV. • Interference between the resonance and potential scattering amplitudes. • Epithermal neutron beams used in BNCT

  20. Fabrication and high heat flux test of divertor cooling elements

    International Nuclear Information System (INIS)

    The plasma facing components in ITER are subjected to a high heat flux from fusion plasma. The heat flux is not only higher than that of existing tokamaks but also has a longer pulse duration (burn time). To minimize a peaking of the heat flux, the thermal deformation towards the plasma should be restrained. One-meter-long monoblock divertor modules with a sliding support structure were fabricated and tested at JAERI. Two kinds of support mechanisms were provided to minimize the thermal deformation of the modules in the upward and downward directions ; one is a pin type sliding structure and the other is a rail type support structure. Both modules were tested on the electron beam HHF test facility, JEBIS (JAERI Electron Beam Irradiation System), in JAERI. The steady-state heat flux of 15 MW/m2 was applied to the surface of the modules to simulate the design condition of ITER CDA. As a result of the HHF test, the performance of the sliding support structures was successfully demonstrated. Three dimensional elastic stress analyses were conducted using a finite element method. The result shows that the relatively high thermal stress is observed at the cooling tube ; and that the maximum thermal stress at the cooling tube exceeds its yield strength. It is necessary to perform the lifetime evaluation of the copper cooling tube against cyclic thermal stresses. (author)

  1. Solid-State Neutron Flux Monitoring Instruments for Nuclear Reactors

    International Nuclear Information System (INIS)

    This paper describes solid-state picoammeters, log-N amplifiers and period meters which have been developed for the flux monitoring and control system of the material testing reactor (JMTR). Recent developments and improvements of insulated-gate field-effect transistors (MOS FET's) have enabled us to realize a perfectly transistorized direct coupled electrometer. The examination of the behaviour of the MOS FET for the ambient temperature variation shows that the voltage drift referred to the gate is higher than that of the ordinary FET, but low enough for picoammeter application. For log-N amplifier applications, selection of the first stage FET or adjustment of the drain current for the individual FET is necessary. The paper also describes the method of compensating the temperature dependence of the log-diode. Balancing out the variation due to the temperature dependent saturation current with an identical diode and compensating the slope by the use of an amplifier having a temperature-dependent gain, a variation of less than 0.1 decades is attained in the range of 3 x 10-12 to 3 x 10-4 A for an ambient temperature variation of 25°C. The authors discuss the non-linear representation of the period-indication to compromise requirements of the expanded scale for the convenience of operation and of compressed scale necessary for period trip. Finally, a description is given of complete picoammeter circuits covering the ranges of 10-10 to 10-4A, the 8-decade log-N amplifier covering 3 x 10-12 to 3 x 10-4 A, and the period meter. (author)

  2. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. 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

    International Nuclear Information System (INIS)

    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/cm2, 10000C cladding temperature, and (2) 40 h at 40 W/cm2, 12000C 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 13700C

  4. UCN sources at external beams of thermal neutrons. An example of PIK reactor

    CERN Document Server

    Lychagin, E V; Muzychka, A Yu; Nekhaev, G V; Nesvizhevsky, V V; Onegin, M S; Sharapov, E I; Strelkov, A V

    2015-01-01

    We consider ultracold neutron (UCN) sources based on a new method of UCN production in superfluid helium (4He). The PIK reactor is chosen as a perspective example of the application of this idea, which consists of installing a 4He UCN source in a beam of thermal or cold neutrons and surrounding the source with a moderator-reflector, which plays the role of a source of cold neutrons (CNs) feeding the UCN source. The CN flux in the source can be several times larger than the incident flux, due to multiple neutron reflections from the moderator-reflector. We show that such a source at the PIK reactor would provide an order of magnitude larger density and production rate than an analogous source at the ILL reactor. We estimate parameters of a 4He source with solid methane (CH4) or/and liquid deuterium (D2) moderator-reflector. We show that such a source with CH4 moderator-reflector at the PIK reactor would provide the UCN density of ~1x10^5 1/cm^3, and the UCN production rate of ~2x10^7 1/s. These values are resp...

  5. Study of a multi-beam accelerator driven thorium reactor

    International Nuclear Information System (INIS)

    The primary advantages that accelerator driven systems have over critical reactors are: (1) Greater flexibility regarding the composition and placement of fissile, fertile, or fission product waste within the blanket surrounding the target, and (2) Potentially enhanced safety brought about by operating at a sufficiently low value of the multiplication factor to preclude reactivity induced events. The control of the power production can be achieved by vary the accelerator beam current. Furthermore, once the beam is shut off the system shuts down. The primary difference between the operation of an accelerator driven system and a critical system is the issue of beam interruptions of the accelerator. These beam interruptions impose thermo-mechanical loads on the fuel and mechanical components not found in critical systems. Studies have been performed to estimate an acceptable number of trips, and the value is significantly less stringent than had been previously estimated. The number of acceptable beam interruptions is a function of the length of the interruption and the mission of the system. Thus, for demonstration type systems and interruption durations of 1sec 5mins 2500/yr and 50/yr are deemed acceptable. However, for industrial scale power generation without energy storage type systems and interruption durations of t 5mins, the acceptable number of interruptions are 25000, 2500, 250, and 3 respectively. However, it has also been concluded that further development is required to reduce the number of trips. It is with this in mind that the following study was undertaken. The primary focus of this study will be the merit of a multi-beam target system, which allows for multiple spallation sources within the target/blanket assembly. In this manner it is possible to ameliorate the effects of sudden accelerator beam interruption on the surrounding reactor, since the remaining beams will still be supplying source neutrons. The proton beam will be assumed to have an

  6. Simulation of Particle Fluxes at the DESY-II Test Beam Facility

    International Nuclear Information System (INIS)

    In the course of this Master's thesis ''Simulation of Particle Fluxes at the DESY-II Test Beam Facility'' the test beam generation for the DESY test beam line was studied in detail and simulated with the simulation software SLIC. SLIC uses the Geant4 toolkit for realistic Monte Carlo simulations of particles passing through detector material.After discussing the physics processes relevant for the test beam generation and the principles of the beam generation itself, the software used is introduced together with a description of the functionality of the Geant4 Monte Carlo simulation. The simulation of the test beam line follows the sequence of the test beam generation. Therefore, it starts with the simulation of the beam bunch of the synchrotron accelerator DESY-II, and proceeds step by step with the single test beam line components. An additional benefit of this thesis is the provision of particle flux and trajectory maps, which make fluxes directly visible by following the particle tracks through the simulated beam line. These maps allow us to see each of the test beam line components, because flux rates and directions change rapidly at these points. They will also guide the decision for placements of future test beam line components and measurement equipment.In the end, the beam energy and its spread, and the beam rate of the final test beam in the test beam area were studied in the simulation, so that the results can be compared to the measured beam parameters. The test beam simulation of this Master's thesis will serve as a key input for future test beam line improvements.

  7. In situ spatial-profile monitoring of beam flux of neutral free radicals produced by photo-deionization of negative ion beams

    International Nuclear Information System (INIS)

    Ion-current difference measurement by light intensity modulation (ICD) is introduced as a convenient method to characterize a purified beam of momentum-controlled neutral free radicals produced by photo-deionization of a negative ion beam for the purpose of surface-reaction-selective device processing. The ICD setup developed in this study to estimate the number flux of the photo-deionized neutral particles exhibited the high precision, sensitivity, and spatial resolution.

  8. Progress of Filtered Neutron Beams Development and Applications at the Horizontal Channels No.2 and No.4 of Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    The neutron filter technique has been applied to create mono-energetic neutron beams with high intensity, at the horizontal channels No.2 and No.4 of the Dalat nuclear research reactor. The mono-energetic neutron beams that have been developed for researches and applications are thermal (0.025 eV), 24 keV, 54 keV, 59 keV, 133 keV and 148 keV. The relative intensities of main peak in filtered neutron energy spectra and the collimated neutron fluxes at the sample irradiation positions are 90 - 96% and 2.8×105 - 7.8×106 n/cm2.s, respectively. Monte Carlo simulations and transmission calculations were performed to each neutron energy beam for optimal design of geometrical structure and neutron filter materials. These filtered neutron beams have been applied efficiently for experimental researches on neutron total and capture cross sections measurements, and elemental analysis in various kinds of samples based on the prompt gamma neutron activation analysis method. This paper reviews the progress of filtered neutron beams development and its applications for past many years at the Dalat nuclear research reactor. (author)

  9. Applicability of copper alloys for DEMO high heat flux components

    Science.gov (United States)

    Zinkle, Steven J.

    2016-02-01

    The current state of knowledge of the mechanical and thermal properties of high-strength, high conductivity Cu alloys relevant for fusion energy high heat flux applications is reviewed, including effects of thermomechanical and joining processes and neutron irradiation on precipitation- or dispersion-strengthened CuCrZr, Cu-Al2O3, CuNiBe, CuNiSiCr and CuCrNb (GRCop-84). The prospects for designing improved versions of wrought copper alloys and for utilizing advanced fabrication processes such as additive manufacturing based on electron beam and laser consolidation methods are discussed. The importance of developing improved structural materials design criteria is also noted.

  10. Heavy ion beam transport through liquid lithium first wall ICF reactor cavities

    International Nuclear Information System (INIS)

    This analysis addresses the critical issue of the final transport of a heavy ion beam in an inertial confinement fusion reactor. The beam must traverse the reaction chamber from the final focusing lens to the target without being disrupted. This requirement has a strong impact on the reactor design. It is essential to the development of ICF fusion reactor technology, that the restrictions placed on the reactor engineering parameters by final beam transport consideration be understood early on

  11. LIBRA - a light ion beam fusion conceptual reactor design

    International Nuclear Information System (INIS)

    The LIBRA light ion beam fusion commercial reactor study is a self-consistent conceptual design of a 330 MWe power plant with an accompanying economic analysis. Fusion targets are imploded by 4 MJ shaped pulses of 30 MeV Li ions at a rate of 3 Hz. The target gain is 80, leading to a yield of 320 MJ. The high intensity part of the ion pulse is delivered by 16 diodes through 16 separate z-pinch plasma channels formed in 100 torr of helium with trace amounts of lithium. The blanket is an array of porous flexible silicon carbind tubes with Li17Pb83 flowing downward through them. These tubes (INPORT units) shield the target chamber wall from both neutron damage and the shock overpressure of the target explosion. The target chamber is 'self-pumped' by the target explosion generated overpressure into a surge tank partially filled with Li17Pb83 that surrounds the target chamber. This scheme refreshes the chamber at the desired 3 Hz frequently without excessive pumping demands. The blanket multiplication is 1.2 and the tritium breeding ratio is 1.4. The direct capital cost of a 331 MWe LIBRA design is estimated to be 2843 Dollar/kWe while a 1200 MWe LIBRA design will cost approximately 1300 Dollar/kWe. (orig.)

  12. Reliable estimation of neutron flux in BWR reactor vessel using the tort code (2) application to neutron and gamma flux estimation

    International Nuclear Information System (INIS)

    A neutron and gamma flux distribution around the core of BWR commercial plant in Japan was calculated, using a three-dimensional transport code, TORT in DOORS32 code system. In the external of the core, the bottom of the model was at an elevation of 150 cm below the bottom of active fuel, the top of the model was at an elevation of the top of the shroud head dome and the radial part of the model was to the outside of the reactor pressure vessel. The top guide beams were modeled explicitly to obtain the neutron and gamma flux distribution both in the beams and outside beams. The each control rod guide tube was also modeled with homogeneous region which included the blade wing and poison tubes so that we could obtain the neutron and gamma flux distribution around the each control rod guide tube. The calculation model mentioned above needed very large memory size which exceeded a few decade giga-bytes. As the using the splicing/coupling method had uncertainly at the splicing/coupling boundary, in this work the calculation was performed without this splicing/coupling method. On the other hand, radioactivity data were measured for a few pieces of the top guide beam, shroud and in-core monitor guide tube in the same plant which was analyzed in the above calculation. So the calculation results were able to be compared with those measured data as benchmarking and at the end of this task, the C/M values at these measured points were obtained and calculation model using TORT was evaluated. (authors)

  13. Gamma and Neutron Flux of a Prompt Gamma Neutron Activation Analysis Collimator at the PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    A Prompt Gamma Neutron Activation Analysis (PGNAA) facility is being studied for installation at PUSPATI TRIGA Reactor (RTP) under the Thorium Flagship programme. This work presents the preliminary design of a PGNAA collimator at the RTP. The result of calculations for gamma and neutron flux at various positions of the PGNAA collimator in the RTP beam port 1 by using the computer code MCNPX are presented and discussed. The results indicate the technical feasibility of the installation of PGNAA facility at the RTP and the possibility of enhancing the utilization of the RTP. (author)

  14. Transport calculation of neutron flux distribution in reflector of PW reactor

    International Nuclear Information System (INIS)

    Two-dimensional transport calculation of the neutron flux and spectrum in the equatorial plain of PW reactor, using computer program DOT 3, is presented. Results show significant differences between neutron fields in which test samples and reactor vessel are exposed. (author)

  15. Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated

  16. Framework for a sustainable development of neutron beam work in the smaller research reactors

    International Nuclear Information System (INIS)

    The authors analyze the present situation of research reactors for neutron beam work in the light of the changes that took place in the nuclear field during the last decades. Trends in supply and demand of neutron beam time in view of the specific requirements of the techniques and of the user's community are outlined. It is argued that resources, both human and material, should be considered in a global perspective, encompassing the national, regional and international levels, where national facilities, mostly low flux research reactors, should be looked upon as a valuable component of a commonwealth of resources to be usefully exploited for the benefit of the neutron user's community at large. The importance of international cooperation to develop a higher level of research reactor utilization is emphasized while suggestions concerning the role of IAEA are made, particularly, to promote the mobility of scientists and engineers directed from developed to less developed countries (LDC's) where research reactors are in operation. The potential of small research reactors in LDC's as an instrument of the country's general scientific and technological development is pointed out as well as difficulties commonly experienced and essential requirements of a successful performance with emphasis on the importance of establishing close links with the national scientific community and especially with university groups. The scientific and technological relevance of neutron scattering techniques is discussed. Reference is made to the techniques best suited to modest research reactor facilities as well as to the importance of developing a local competence in instrument design, optimization and construction. (author). 12 refs

  17. High energy ion beam mixing

    International Nuclear Information System (INIS)

    Experimental investigations have been made on the parameters which can be used to control the mixing profiles, and the width of intermixed layers in film-substrate systems being irradiated by high energy heavy ion beams. The samples were irradiated by ion beams of Au, Cu, and Si with energies of 1.5 to 3 MeV. Typical examples of the RBS spectra are presented and discussions are made on the extent of contribution of binary collisions on the interfacial mixing. The experimental and simulation results show that the interfacial mixing is dominated by the binary collisions. (author)

  18. A high-performance electron beam ion source

    Energy Technology Data Exchange (ETDEWEB)

    Alessi,J.; Beebe, E.; Bellavia, S.; Gould, O.; Kponou, A.; Lambiase, R.; Lockey, R.; McCafferty, D.; Okamura, M.; Pikin, A. I.; Raparia, D.; Ritter, J.; Syndstrup, L.

    2009-06-08

    At Brookhaven National Laboratory, a high current Electron Beam Ion Source (EBIS) has been developed as part of a new preinjector that is under construction to replace the Tandem Van de Graaffs as the heavy ion preinjector for the RHIC and NASA experimental programs. This preinjector will produce milliampere-level currents of essentially any ion species, with q/A {ge} 1/6, in short pulses, for injection into the Booster synchrotron. In order to produce the required intensities, this EBIS uses a 10A electron gun, and an electron collector designed to handle 300 kW of pulsed electron beam power. The EBIS trap region is 1.5 m long, inside a 5T, 2m long, 8-inch bore superconducting solenoid. The source is designed to switch ion species on a pulse-to-pulse basis, at a 5 Hz repetition rate. Singly-charged ions of the appropriate species, produced external to the EBIS, are injected into the trap and confined until the desired charge state is reached via stepwise ionization by the electron beam. Ions are then extracted and matched into an RFQ, followed by a short IH Linac, for acceleration to 2 MeV/A, prior to injection into the Booster synchrotron. An overview of the preinjector is presented, along with experimental results from the prototype EBIS, where all essential requirements have already been demonstrated. Design features and status of construction of the final high intensity EBIS is also be presented.

  19. Neutronic modeling and thermal neutron flux measurement of the MCR6 rod in the NRU reactor

    International Nuclear Information System (INIS)

    The six-barrel Multiple-Capsule Water Cooled Rods (MCR6) in the NRU reactor have been used to produce isotopes such as I-131 and Ir-192. This paper describes the modeling of the MCR6 rods and the simulation method used to predict the neutron fluxes. The sensitivity of various radioisotope loadings of MCR6 rods upon flux and power perturbations of neighbouring rods is investigated. This paper also presents the results of thermal neutron flux measurements in one of the MCR6 rods from gold wire detectors using the neutron activation technique. The measured fluxes match very well with the simulated fluxes, within ±2%. (author)

  20. Neutronic modeling and thermal neutron flux measurement of the MCR6 rod in the NRU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, X.; Leung, T.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2009-07-01

    The six-barrel Multiple-Capsule Water Cooled Rods (MCR6) in the NRU reactor have been used to produce isotopes such as I-131 and Ir-192. This paper describes the modeling of the MCR6 rods and the simulation method used to predict the neutron fluxes. The sensitivity of various radioisotope loadings of MCR6 rods upon flux and power perturbations of neighbouring rods is investigated. This paper also presents the results of thermal neutron flux measurements in one of the MCR6 rods from gold wire detectors using the neutron activation technique. The measured fluxes match very well with the simulated fluxes, within {+-}2%. (author)

  1. Software for determination of the thermal neutron flux in a nuclear reactor

    International Nuclear Information System (INIS)

    In this study thermal neutron flux distribution was performed using the activity of the irradiated activation detectors: Au and In foils. The neutron flux in the IEA-R1 nuclear reactor at IPEN/SP in the pneumatic station was then house software and the results were compared. The data permits a discussion about the performed of this software. (author)

  2. Chopping high intensity proton beams

    Energy Technology Data Exchange (ETDEWEB)

    Wiesner, Christoph; Dinter, Hannes; Droba, Martin; Meusel, Oliver; Mueller, Ilja; Noll, Daniel; Payir, Onur; Ratzinger, Ulrich; Schneider, Philipp [IAP, Goethe-Universitaet Frankfurt, Max-von-Laue-Str. 1, 60438 Frankfurt am Main (Germany)

    2013-07-01

    A novel E x B chopper system for high intensity proton beams is being developed to deliver 100 ns beam pulses in the low energy transport line of the accelerator driven neutron source FRANZ. It combines a static magnetic deflection field with a pulsed electric compensation field in a Wien filter-type E x B configuration. Behind the deflection unit a massless septum system is used for beam separation. The setup minimizes the risk of voltage breakdowns and provides secure beam dumping outside the transport line. The electric deflection field is driven by a HV pulse generator providing ±6 kV at a repetition rate of 250 kHz. Accurate layout of the deflection plates is required to tackle the issues of field quality, cooling and spark prevention. Careful matching of electric and magnetic deflection forces is required to prevent aberrations and emittance growth. Numerical studies for the field design and their effects on beam transport are presented and an overview of the hardware development is given.

  3. The conceptual calculation for the neutron beam device at Mark 1

    International Nuclear Information System (INIS)

    The thermal neutron beam device, epithermal neutron beam device and test duct experiment device are designed by using Monte Carlo method at 30 kW Mark 1(-1). The compared calculation for transverse cross section dimension, moderator, reflector and others of neutron filter device are studied in this paper. The three optimized neutron beams including thermal neutron beam, epithermal neutron beam and the beam for measuring blood boron density, whose neutron flux density per reactor power are rather high, are also introduced. The results show that the BNCT neutron beam can be designed by using 30kW -1 reactor. (author)

  4. On the utilization of neutron beams of research reactors in research and applications

    International Nuclear Information System (INIS)

    Nuclear research reactors are the most widely available neutron sources, and they are capable of producing very high fluxes of neutrons having a considerable range of energies, from a few MeV to 10 MeV. Therefore, these neutrons can be used in many fields of basic research and for applications in physics, chemistry, medicine, biology, etc. Experiments with research reactors over the last 50 years have laid the foundations of today's nuclear technology. In addition, research reactors continue to be utilized as facilities for testing materials and in training manpower for nuclear programs, because basic training on a research reactor provides an essential understanding of the nuclear process, and personnel become accustomed to work under the special conditions resulting from irradiation and contamination risks

  5. Slow neutron flux extrapolation distances in R-5 and CIRUS reactors

    International Nuclear Information System (INIS)

    In order to calculate the core reactivity, fuel channel power outputs and neutron flux levels in the R-5 reactor at Trombay, axial flux extrapolation distances are required. For this, an analysis is carried out considering the reactor core as a two region neutron multiplying system in axial direction. The slow neutron diffusion equations for both the regions are solved analytically by applying suitable boundary conditions. Application of this method for the estimation of top extrapolation distances in CIRUS, has given results which agree well with accepted values for the reactor. (author)

  6. High flux lithium antineutrino source with variable hard spectrum

    CERN Document Server

    Lyashuk, V I

    2016-01-01

    The high flux antineutrino source with hard antineutrino spectrum based on neutron activation of 7Li and subsequent fast beta-decay (T 1/2 = 0.84 s) of the 8Li isotope with emission of antineutrino with energy up to 13 MeV - is discussed. Creation of the intensive isotope neutrino source of hard spectrum will allow to increase the detection statistics of neutrino interaction and it is especially urgent for oscillation experiments. The scheme of the proposed neutrino source is based on the continuous transport of the created 8Li to the neutrino detector, which moved away from the place of neutron activation. Analytical expressions for lithium antineutrino flux is obtained. The discussed source will ensure to increase the cross section for reactions with deuteron from several times to tens compare to the reactor antineutrino spectrum. An another unique feature of the installation is the possibility to vary smoothly the hardness of the antineutrino spectrum.

  7. Role of plasma enhanced atomic layer deposition reactor wall conditions on radical and ion substrate fluxes

    Energy Technology Data Exchange (ETDEWEB)

    Sowa, Mark J., E-mail: msowa@ultratech.com [Ultratech/Cambridge NanoTech, 130 Turner Street, Building 2, Waltham, Massachusetts 02453 (United States)

    2014-01-15

    Chamber wall conditions, such as wall temperature and film deposits, have long been known to influence plasma source performance on thin film processing equipment. Plasma physical characteristics depend on conductive/insulating properties of chamber walls. Radical fluxes depend on plasma characteristics as well as wall recombination rates, which can be wall material and temperature dependent. Variations in substrate delivery of plasma generated species (radicals, ions, etc.) impact the resulting etch or deposition process resulting in process drift. Plasma enhanced atomic layer deposition is known to depend strongly on substrate radical flux, but film properties can be influenced by other plasma generated phenomena, such as ion bombardment. In this paper, the chamber wall conditions on a plasma enhanced atomic layer deposition process are investigated. The downstream oxygen radical and ion fluxes from an inductively coupled plasma source are indirectly monitored in temperature controlled (25–190 °C) stainless steel and quartz reactors over a range of oxygen flow rates. Etch rates of a photoresist coated quartz crystal microbalance are used to study the oxygen radical flux dependence on reactor characteristics. Plasma density estimates from Langmuir probe ion saturation current measurements are used to study the ion flux dependence on reactor characteristics. Reactor temperature was not found to impact radical and ion fluxes substantially. Radical and ion fluxes were higher for quartz walls compared to stainless steel walls over all oxygen flow rates considered. The radical flux to ion flux ratio is likely to be a critical parameter for the deposition of consistent film properties. Reactor wall material, gas flow rate/pressure, and distance from the plasma source all impact the radical to ion flux ratio. These results indicate maintaining chamber wall conditions will be important for delivering consistent results from plasma enhanced atomic layer deposition

  8. 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

    Energy Technology Data Exchange (ETDEWEB)

    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%.

  9. 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

    International Nuclear Information System (INIS)

    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 3000C (in HFIR). Precipitation of MC was observed in 20% cold-worked JPCA. Swelling of all specimens was less than 1%

  10. Liquid metal target studies for high-power proton beams

    International Nuclear Information System (INIS)

    Full text: Proton beams with an average beam power of several MW are needed to produce the high secondary particle flux required by future accelerator centres like neutrino factories or neutron spallation sources. In such scenarios, the secondary beams are produced via conversion targets and the high power involved imposes the need for advanced target concepts. The use of liquid metal is a natural solution to the stresses and fatigue, induced by the proton beam, that eventually lead to the destruction of most solid targets. A liquid jet would provide a 'new' target for each proton pulse if the material disrupted by the proton beam can be evacuated within the proton pulse interval. Such a target configuration is presently considered for the pion production target of a neutrino superbeam or a neutrino factory. We present experimental results on the impact of proton beam pulses on free-surface liquid metal targets. These tests were performed at the ISOLDE facility at CERN. (author)

  11. Fast flux fluid fuel reactor: A concept for the next generation of nuclear power production

    International Nuclear Information System (INIS)

    Nuclear energy has not become the preferred method of electrical energy production largely because of economic, safety, and proliferation concerns and challenges posed by nuclear waste disposal. Economies is the most important factor. To reduce the capital costs, the authors propose a compact configuration with a very high power density and correspondingly reduced reactor component sizes. Enhanced efficiency made possible by higher operating temperatures will also improve the economics of the design, and design simplicity will keep capital, operational, and maintenance costs down. The most direct solution to the nuclear waste problem is to eliminate waste production or, at least, minimize its amount and long-term radiotoxicity. This can be achieved by very high burnups, ideally 100%, and by the eventual transmutation of the long-lived fission products in situ. Very high burnups also improve the economics by optimal exploitation of the fuel. Safety concerns can be addressed by an inherently safe reactor design. Because of the intrinsic nature of nuclear materials, there probably is no definitive answer to proliferation concerns for systems that generate neutrons; however, it is important to minimize proliferation risks. The thorium cycle is a promising option because (a) plutonium is produced only in very small quantities, (b) the presence of 232U makes handling the fuel very difficult and therefore proliferation resistant, and (c) 233U is a fissile isotope that is less suitable than 239Pu for making weapons and can be diluted with other uranium isotopes. An additional benefit of the thorium cycle is that it increases nuclear fuel resources by one order of magnitude. A fast flux fluid fuel reactor is a concept that can satisfy all the foregoing requirements. The fluid fuel systems have a very simple structure. Because integrity of the fuel is not an issue, these systems can operate at very high temperatures, can have high power densities, and can achieve very high

  12. Multipurpose Utilisation of a Medium Flux Research Reactor. Benefit for the Society

    International Nuclear Information System (INIS)

    The Budapest Research Reactor (BRR) was restarted after a major refurbishment and increase in power to 10 MW in 1992. Basically, the experience gained with the utilization of this multipurpose facility during the past 20 years is described here. The utilization aims for 3 major activities: i) Research and development base for the energy sector: scientific and safety support for the Paks NPP; research in energy saving and production. ii) A complex source of irradiations for materials testing and modification, diagnostics in nanotechnologies, engineering, healthcare etc. iii) Neutron beams from the horizontal channels of the reactor serve for exploratory as well as for applied research in a very wide range of disciplines. Graduate and professional training is also in the scope of our activity. The reactor went critical first in 1959. It served nearly 3 decades as a home base for learning nuclear sciences and technologies, to development nuclear energetics, which resulted in launching four power plant blocks in the eighties, as well as to establish neutron beam research in our country. Nearly 20 years passed now that the decision was made - after the falling of the Iron Curtain'' - the practically brand new 10 megawatt reactor should be commissioned and opened for the international user community. The reactor reached its nominal power in May 1993 and neutron beam experiments were available on 4 instruments at that time. Thanks to a continuous development the number of experimental stations now is 15, the research staffs has grown from 10 to nearly 50 scientists and research facilities have been improved considerably. A few important milestones should be mentioned: a liquid hydrogen cold source was installed and the neutron guide system was replaced by a supermirror guide configuration, yielding a factor of 50-80 gain in neutron intensity; a second guide hall was constructed to house a new time-of-flight instrument; BRR became a member of the European neutron

  13. Sensibility studies of the equivalent thermal neutron flux on the heat exchanger of a sodium cooled fast reactor. (1. Pt.)

    International Nuclear Information System (INIS)

    This paper reports on sensibility studies of the equivalent thermal neutron flux on the heat exchanger for a sodium cooled fast reactor. Graphs and diagrams of the neutron flux in function of the reactor geometry, contribution of the fission sources in the core and the blanket of the reactor are given

  14. Beam removal block and shielding resign for the MARS neutron therapy reactor

    International Nuclear Information System (INIS)

    The beam removal block and shielding design for the MARS neutron therapy reactor are described. The requirements to the beams' characteristics, filters, collimator and reactor shielding are formulated. Radiation field levels in medical box are analyzed for beams' different operation conditions. It is stated that the removal block and shutter compositions meet necessary conditions in radiation treatment and emergency evacuation

  15. Ion beam steering with a high intensity electron beam

    International Nuclear Information System (INIS)

    In conventional theory, steering or bending an ion beam of high energy and high current requires very intense magnetic fields, which are both uneconomical and bulky. This problem is even more severe for a singly charged ion beam with very high atomic number, which requires large magnetic field energy both to bend and also to focus the beam against its self electric field. In this paper we present a new and simple technique, which will substantially alleviate these problems

  16. Measurement and calculation of the neutron flux distribution in the RP-10 reactor

    International Nuclear Information System (INIS)

    In this work implementing experimental methods are implemented for easy reproduction for measuring the spatial distribution or thermal neutron flux in the RP-10 reactor core. Using two measuring methods: the passive and the active ones. In the passive method was used the activation technique using foils such as gold, manganese, and indium. These were irradiated in the reactor core and treated through the Westcott's formalism. In the active method was used the Self Powered Neutron Detectors (SPNs) for which was necessary to condition the detectors response for the data acquisition. The knowledge of the spatial distribution of RP-10 reactor neutrons flux will contribute in the understanding of other interesting parameters of reactor physics such as power density, reactivity, buckling, etc.. Wish knowledge is important for reactor operation. Fuel burnup calculations as well as others related to safety. (author)

  17. High current beam transport with multiple beam arrays

    International Nuclear Information System (INIS)

    Highlights of recent experimental and theoretical research progress on the high current beam transport of single and multiple beams by the Heavy Ion Fusion Accelerator Research (HIFAR) group at the Lawrence Berkeley Laboratory (LBL) are presented. In the single beam transport experiment (SBTE), stability boundaries and the emittance growth of a space charge dominated beam in a long quadrupole transport channel were measured and compared with theory and computer simulations. Also, a multiple beam ion induction linac (MBE-4) is being constructed at LBL which will permit study of multiple beam transport arrays, and acceleration and bunch length compression of individually focused beamlets. Various design considerations of MBE-4 regarding scaling laws, nonlinear effects, misalignments, and transverse and longitudinal space charge effects are summarized. Some aspects of longitudinal beam dynamics including schemes to generate the accelerating voltage waveforms and to amplify beam current are also discussed

  18. Brookhaven - going for gold

    International Nuclear Information System (INIS)

    Experiments using polarized (spin oriented) protons have a promising future at the Laboratory. Polarized sources are being improved considerably, and the advent of the Booster will enable accumulation of polarized beams before injection into the main ring. Adding these factors together indicates the possibility of colliding polarized proton beams in the proposed RHIC collider (of which more later) with a luminosity as high as 1032 and an energy of 350 GeV. The newcomer to the programme at the AGS is the series of experiments with light ion beams drawn from the nearby tandem. These began last October with oxygen-8 and silicon-14 beams at energies up to 15 GeV per nucleon and 5x108 accelerated ions per pulse for oxygen. (orig./HSI).

  19. Boron neutron capture therapy (BNCT) for glioblastoma multiforme using the epithermal neutron beam at the Brookhaven Medical Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Capala, J. [Brookhaven National Lab., Upton, NY (United States); Diaz, A.Z.; Chadha, M. [Univ. Hospital, State Univ. of New York, NY (United States)] [and others

    1997-12-31

    The abstract describes evaluation of boron neutron capture therapy (BNCT) for two groups of glioblastoma multiforme patients. From September 1994 to February 1996 15 patients have been treated. In September 1997 another 34 patients were examined. Authors determined a safe starting dose for BNCT using epithermal neutrons and BPA-F. They have also evaluated adverse effects of BNCT at this starting dose. Therapeutic effectiveness of this starting dose has been evaluated. No significant side effects from BPA-F infusion or BNCT treatment were observed in normal brains.

  20. Design of the experimental apparatus to obtain a thermal neutron beam, intermediate-energy neutrons (2-144 keV) and high-energy photons (6 MeV) by means of the TRIGA reactor at the ENEA Casaccia center

    CERN Document Server

    Laitano, R F

    1987-01-01

    Design of the experimental apparatus to obtain a thermal neutron beam, intermediate-energy neutrons (2-144 keV) and high-energy photons (6 MeV) by means of the TRIGA reactor at the ENEA Casaccia center

  1. Flux measurements in a nuclear research reactor by using an aluminum nitride detector

    International Nuclear Information System (INIS)

    A small polycrystalline aluminium nitride detector with a thickness of 381 μm was used to measure a 200,000 Ci Co60 source and to measure the flux in a research reactor where the neutron flux is about 1014/cm2 s, which is nearly the same order as in the commercial power plant. If the applied voltage is greater than or equal to 2000 V and if the measurements are done in a short period of time so that the heat energy does not build up in the aluminium nitride, then the measured electric current is linearly proportional to the input flux. It is assumed of course that the energy spectrum of the input flux remains constant. This linearity relation is illustrated by the results of a measurement in which the reactor power has been controlled so that the flux becomes a step function

  2. Reference equilibrium core with central flux irradiation facility for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    In order to assess various core parameters a reference equilibrium core with Low Enriched Uranium (LEU) fuel for Pakistan Research Reactor (PARR-1) was assembled. Due to increased volume of reference core, the average neutron flux reduced as compared to the first higher power operation. To get a higher neutron flux an irradiation facility was created in centre of the reference equilibrium core where the advantage of the neutron flux peaking was taken. Various low power experiments were performed in order to evaluate control rods worth and neutron flux mapping inside the core. The neutron flux inside the central irradiation facility almost doubled. With this arrangement reactor operation time was cut down from 72 hours to 48 hours for the production of the required specific radioactivity. (author)

  3. Prompt gamma activation analysis using mobile reactor neutron beam

    International Nuclear Information System (INIS)

    Among the nuclear analytical methods that have proved very useful in biological and medical analyses is the in vivo prompt gamma neutron activation analysis (IVPGAA). In this work, an IVPGAA facility was assembled on a zero-power mobile nuclear reactor and has demonstrated its versatility for in vivo medical diagnosis. Absolute measurements of some environmental contaminants such as Cd, Hg, and Si in organs can be determined rapidly by partial body scan of IVPGAA, while assessment of vital constituents such as Ca, Cl, N, and P in either whole body or body part can be scanned by IVPGAA technique effectively. The in vivo clinical application using mobile reactor neutron beam are reviewed in detail. The IVPGAA scan provides unique insight into elemental concentration purpose. The IVPGAA scan can be performed on a regular basis without discomfort and radiation risk for patients. (author)

  4. Particles fluidized bed receiver/reactor with a beam-down solar concentrating optics: 30-kWth performance test using a big sun-simulator

    Science.gov (United States)

    Kodama, Tatsuya; Gokon, Nobuyuki; Cho, Hyun Seok; Matsubara, Koji; Etori, Tetsuro; Takeuchi, Akane; Yokota, Shin-nosuke; Ito, Sumie

    2016-05-01

    A novel concept of particles fluidized bed receiver/reactor with a beam-down solar concentrating optics was performed using a 30-kWth window type receiver by a big sun-simulator. A fluidized bed of quartz sand particles was created by passing air from the bottom distributor of the receiver, and about 30 kWth of high flux visible light from 19 xenon-arc lamps of the sun-simulator was directly irradiated on the top of the fluidized bed in the receiver through a quartz window. The particle bed temperature at the center position of the fluidized bed went up to a temperature range from 1050 to 1200°C by the visible light irradiation with the average heat flux of about 950 kW/m2, depending on the air flow rate. The output air temperature from the receiver reached 1000 - 1060°C.

  5. High performance light water reactor

    International Nuclear Information System (INIS)

    The objective of the high performance light water reactor (HPLWR) project is to assess the merit and economic feasibility of a high efficiency LWR operating at thermodynamically supercritical regime. An efficiency of approximately 44% is expected. To accomplish this objective, a highly qualified team of European research institutes and industrial partners together with the University of Tokyo is assessing the major issues pertaining to a new reactor concept, under the co-sponsorship of the European Commission. The assessment has emphasized the recent advancement achieved in this area by Japan. Additionally, it accounts for advanced European reactor design requirements, recent improvements, practical design aspects, availability of plant components and the availability of high temperature materials. The final objective of this project is to reach a conclusion on the potential of the HPLWR to help sustain the nuclear option, by supplying competitively priced electricity, as well as to continue the nuclear competence in LWR technology. The following is a brief summary of the main project achievements:-A state-of-the-art review of supercritical water-cooled reactors has been performed for the HPLWR project.-Extensive studies have been performed in the last 10 years by the University of Tokyo. Therefore, a 'reference design', developed by the University of Tokyo, was selected in order to assess the available technological tools (i.e. computer codes, analyses, advanced materials, water chemistry, etc.). Design data and results of the analysis were supplied by the University of Tokyo. A benchmark problem, based on the 'reference design' was defined for neutronics calculations and several partners of the HPLWR project carried out independent analyses. The results of these analyses, which in addition help to 'calibrate' the codes, have guided the assessment of the core and the design of an improved HPLWR fuel assembly. Preliminary selection was made for the HPLWR scale

  6. Measurement of the thermal flux distribution in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    The knowledge of the neutron flux distribution in research reactors is important because it gives the power distribution over the core, and it provides better conditions to perform experiments and sample irradiations. The measured neutron flux distribution can also be of interest as a means of comparison for the calculational methods of reactor analysis currently in use at this institute. The thermal neutron flux distribution of the IEA-R1 reactor has been measured with the miniature chamber WL-23292. For carrying out the measurements, it was buit a guide system that permit the insertion of the mini-chamber i between the fuel of the fuel elements. It can be introduced in two diferent positions a fuel element and in each it spans 26 axial positions. With this guide system the thermal neutron flux distribution of the IEA-R1 nuclear reactor can be obtained in a fast and efficient manner. The element measured flux distribution shows clearly the effects of control rods and reflectors in the IEA-R1 reactor. The difficulties encountered during the measurements are mentioned with detail as well as the procedures adopteed to overcome them. (Author)

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

    International Nuclear Information System (INIS)

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

  8. Reactor core design of Gas Turbine High Temperature Reactor 300

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute (JAERI) has been designing Japan's original gas turbine high temperature reactor, Gas Turbine High Temperature Reactor 300 (GTHTR300). The greatly simplified design based on salient features of the High Temperature Gas-cooled Reactor (HTGR) with a closed helium gas turbine enables the GTHTR300 a highly efficient and economically competitive reactor to be deployed in early 2010s. Also, the GTHTR300 fully taking advantage of various experiences accumulated in design, construction and operation of the High Temperature Engineering Test Reactor (HTTR) and existing fossil fired gas turbine systems reduces technological development concerning a reactor system and electric generation system. Original design features of this system are the reactor core design based on a newly proposed refueling scheme named sandwich shuffling, conventional steel material usage for a reactor pressure vessel (RPV), an innovative coolant flow scheme and a horizontally installed gas turbine unit. The GTHTR300 can be continuously operated without the refueling for 2 years. Due to these salient features, the capital cost of the GTHTR300 is less than a target cost of 200,000 yen (1667 US$)/kW e, and the electric generation cost is close to a target cost of 4 yen (3.3 US cents)/kW h. This paper describes the original design features focusing on the reactor core design and the in-core structure design, including the innovative coolant flow scheme for cooling the RPV. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan

  9. Operational characteristics of the high flux plasma generator Magnum-PSI

    International Nuclear Information System (INIS)

    Highlights: •We have described the design and capabilities of the plasma experiment Magnum-PSI. •The plasma conditions are well suited for PSI studies in support of ITER. •Quasi steady state heat fluxes over 10 MW m−2 have been achieved. •Transient heat and particle loads can be generated to simulate ELM instabilities. •Lithium coating can be applied to the surfaces of samples under vacuum. -- Abstract: In Magnum-PSI (MAgnetized plasma Generator and NUMerical modeling for Plasma Surface Interactions), the high density, low temperature plasma of a wall stabilized dc cascaded arc is confined to a magnetized plasma beam by a quasi-steady state axial magnetic field up to 1.3 T. It aims at conditions that enable fundamental studies of plasma–surface interactions in the regime relevant for fusion reactors such as ITER: 1023–1025 m−2 s−1 hydrogen plasma flux densities at 1–5 eV. To study the effects of transient heat loads on a plasma-facing surface, a high power pulsed magnetized arc discharge has been developed. Additionally, the target surface can be transiently heated with a pulsed laser system during plasma exposure. In this contribution, the current status, capabilities and performance of Magnum-PSI are presented

  10. Applications of heat pipes for high thermal load beam lines

    International Nuclear Information System (INIS)

    The high flux beam produced by insertion devices often requires special heat removal techniques. For the optical elements used in such high thermal load beam lines the required precision demands a highly accurate design. Heat pipe cooling of critical elements of the x-1 beam line at the National Synchrotron Light Source is described. This method reduces vibrations caused by water cooling systems and simplifies the design. In some of these designs, deposited heat must be transferred through unbonded contact interfaces. A pinhole assembly and a beam position monitor designed for the x-1 beam line both transfer heat through such interfaces in an ultrahigh vacuum environment. The fundamental design objective is that of removing the heat with minimal interface thermal resistance. We present our test method and results for measuring the thermal resistance across metallic interfaces as a function of contact pressure. The design of some devices which utilize both heat pipes and thermal contact interfaces will also be described. (orig.)

  11. High power neutral beam systems

    International Nuclear Information System (INIS)

    Spurred by the requirement to supply megawatts of power to heat magnetically confined plasmas to temperatures of interest for fusion research, a new class of low energy, high power accelerators termed neutral beam injectors has been developed. Industry has played an important role in building upon technology advances at the national laboratories to engineer neutral beam injectors to meet the needs of specific users. A brief retrospective of the field is presented, with emphasis upon one particular application, that of DIII-D, a large tokamak at General Atomics. In this instance, the role of industry has been especially extensive because the user/system integrator is itself an industrial concern. 4 refs., 7 figs., 2 tabs

  12. Effects of Collisional Dissipation on the "Colliding Beam Fusion Reactor "

    Science.gov (United States)

    Lampe, Martin; Manheimer, Wallace M.

    1998-11-01

    Rostoker, Binderbauer and Monkhorst have recently proposed a "colliding beam fusion reactor" (CBFR) for use with the p-B11 reaction. We have examined the various dissipative processes resulting from Coulomb collisions, and have concluded that the CBFR equilibrium cannot be sustained for long enough to permit net fusion gain. There are many collisional processes which occur considerably faster than fusion, and result in particle loss, energy loss, or detuning of the resonant energy for the p-B reaction. Pitch-angle scattering of protons off the boron beam, which occurs 100 times faster than fusion, isotropizes the proton beam and results in proton loss. Energy exchange between protons and boron, which is 20 times faster than fusion, detunes the resonance. Proton-proton scattering, which is faster than fusion for all CBFR scenarios, Maxwellianizes the protons and thus detunes the resonance. Ion-electron collisions lead indirectly to a friction between the two ion beams, which is typically fast compared to the fusion process. Results of Fokker-Planck analyses of each process will be shown.

  13. High density operation for reactor-relevant power exhaust

    Science.gov (United States)

    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.

  14. High density operation for reactor-relevant power exhaust

    Energy Technology Data Exchange (ETDEWEB)

    Wischmeier, M., E-mail: marco.wischmeier@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, D-85748 Garching (Germany)

    2015-08-15

    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.

  15. Calculation of neutron flux in PUSPATI TRIGA MARK II reactor using Monte-Carlo n-particle approach

    International Nuclear Information System (INIS)

    A Monte Carlo simulation of neutron flux at the TRIGA MARK II PUSPATI (RTP) nuclear research reactor at Agensi Nuklear Malaysia was carried out using the MCNP5 program. The objective of the work is to simulate the neutron flux inside the reactor core. Calculations of neutron flux for fast and thermal neutron were carried out under the conditions in which the control rod was either fully withdrawn from or fully inserted into the reactor. (Author)

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

    Science.gov (United States)

    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.

  17. Use of CMOS imagers to measure high fluxes of charged particles

    Science.gov (United States)

    Servoli, L.; Tucceri, P.

    2016-03-01

    The measurement of high flux charged particle beams, specifically at medical accelerators and with small fields, poses several challenges. In this work we propose a single particle counting method based on CMOS imagers optimized for visible light collection, exploiting their very high spatial segmentation (> 3 106 pixels/cm2) and almost full efficiency detection capability. An algorithm to measure the charged particle flux with a precision of ~ 1% for fluxes up to 40 MHz/cm2 has been developed, using a non-linear calibration algorithm, and several CMOS imagers with different characteristics have been compared to find their limits on flux measurement.

  18. Automatic control of neutron flux in experimental channels of the WWR-M type reactors

    International Nuclear Information System (INIS)

    The flowsheet of the neutron flux local regulator intended for maintaining the given level of neutron flux distribution in experimental channels of the WWR-M type reactor under stationary and transition modes, is suggested. The functional diagram of the electron regulation block (ERB) in considered. The regulator is tested when the reactor operates with the capacity of 13 MWt along with the staff system of automated regulation and without it. The experiments carried out demonstrate the stable operation of the entire control system and good performance characteristics of the ERR block. The conclusion is made that the suggested method of neutron flux automated regulation in experimental channels can be successfully extended to a higher number of experimental channels and applied at other research reactors. Small size fission chambers and direct charging detectors can be used in local systems as sensors

  19. Status of the visible Free-Electron Laser at the Brookhaven Accelerator Test Facility

    International Nuclear Information System (INIS)

    The 500 nm Free-Electron Laser (ATF) of the Brookhaven National Laboratory is reviewed. We present an overview of the ATF, a high-brightness, 50-MeV, electron accelerator and laser complex which is a users' facility for accelerator and beam physics. A number of laser acceleration and FEL experiments are under construction at the ATF. The visible FEL experiment is based on a novel superferric 8.8 mm period undulator. The electron beam parameters, the undulator, the optical resonator, optical and electron beam diagnostics are discussed. The operational status of the experiment is presented. 22 refs., 7 figs

  20. BLIP. [Brookhaven Linac Isotope Producer (BLIP)

    Energy Technology Data Exchange (ETDEWEB)

    Stang, Jr, L G

    1976-01-01

    The operation of the Brookhaven Linac Isotope Producer (BLIP) is discussed. Topics covered include targets, target holders, linac specifications, beam transport, and current production performance. The use of the BLIP is confined exclusively to the development of radionuclides that are, or should be, of medical interest, and the facility is moving rapidly into a self-supporting state from the income of the products. (PMA)