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1

HFIR | High Flux Isotope Reactor | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

HFIR Working with HFIR Neutron imaging offers new tools for exploring artifacts and ancient technology Home | User Facilities | HFIR HFIR | High Flux Isotope Reactor SHARE The High...

2

HIGH FLUX ISOTOPE REACTOR PRELIMINARY DESIGN STUDY  

SciTech Connect

A comparison of possible types of research reactors for the production of transplutonium elements and other isotopes indicates that a flux-trap reactor consisting of a beryllium-reflecteds light-water-cooled annular fuel region surrounding a light-water island provides the required thermal neutron fluxes at minimum cost. The preliminary desigu of such a reactor was carried out on the basis of a parametric study of the effect of dimensions of the island and fuel regions heat removal rates, and fuel loading on the achievable thermal neutmn fluxes in the island and reflector. The results indicate that a 12- to 14-cm- diam. island provides the maximum flux for a given power density. This is in good agreement with the US8R critical experiments. Heat removal calculations indicate that average power densities up to 3.9 Mw/liter are achievable with H/ sub 2/O-cooled, platetype fuel elements if the system is pressurized to 650 psi to prevent surface boiling. On this basis, 100 Mw of heat can be removed from a 14-cm-ID x 36-cm-OD x 30.5-cm-long fuel regions resulting in a thermal neutron flux of 3 x 10/sup 15/ in the island after insertion of 100 g of Cm/sup 244/ or equivalent. The resulting production of Cf/sup 252/ amounts to 65 mg for a 1 1/2- year irradiation. Operation of the reactor at the more conservative level of 67 Mw, providing an irradiation flux of 2 x 10/sup 15/ in the islands will result in the production of 35 mg of Cf/sup 252/ per 18 months from 100 g of Cm/sup 244/. A development program is proposed to answer the question of the feasibility of the higher power operation. In addition to the central irradiation facility for heavyelement productions the HFIR contains ten hydraulic rabbit tubes passing through the beryllium reflector for isotope production and four beam holes for basic research, Preliminary estimates indicate that the cost of the facility, designed for an operating power level of 100 Mw, will be approximately 2 million. (auth)

Lane, J.A.; Cheverton, R.D.; Claiborne, G.C.; Cole, T.E.; Gambill, W.R.; Gill, J.P.; Hilvety, N.; McWherther, J.R.; Vroom, D.W.

1959-03-20T23:59:59.000Z

3

CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Engineering - Oak Ridge National Laboratory High Flux Isotope Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Engineering Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor

4

CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Reactor CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Training Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. RADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor

5

CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Reactor CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Maintenance Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Safety Basis - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor

6

The High Flux Isotope Reactor at Oak Ridge National Laboratory  

NLE Websites -- All DOE Office Websites

The High Flux Isotope Reactor at ORNL The High Flux Isotope Reactor at ORNL Aerial of the High Flux Isotope Reactor Site The High Flux Isotope Reactor site is located on the south side of the ORNL campus and is about a three-minute drive from her sister neutron facility, the Spallation Neutron Source. Operating at 85 MW, HFIR is the highest flux reactor-based source of neutrons for research in the United States, and it provides one of the highest steady-state neutron fluxes of any research reactor in the world. The thermal and cold neutrons produced by HFIR are used to study physics, chemistry, materials science, engineering, and biology. The intense neutron flux, constant power density, and constant-length fuel cycles are used by more than 500 researchers each year for neutron scattering research into

7

High Flux Isotope Reactor (HFIR) | Nuclear Science | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

High Flux Isotope Reactor High Flux Isotope Reactor May 30, 2013 The High Flux Isotope Reactor (HFIR) first achieved criticality on August 25, 1965, and achieved full power in August 1966. It is a versatile 85-MW isotope production, research, and test reactor with the capability and facilities for performing a wide variety of irradiation experiments and a world-class neutron scattering science program. HFIR is a beryllium-reflected, light water-cooled and moderated flux-trap type swimming pool reactor that uses highly enriched uranium-235 as fuel. HFIR typically operates seven 23-to-27 day cycles per year. Irradiation facility capabilities include Flux trap positions: Peak thermal flux of 2.5X1015 n/cm2/s with similar epithermal and fast fluxes (Highest thermal flux available in the

8

CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Management- Oak Ridge National Laboratory High Flux Isotope Management- Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Management in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Management- Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope

9

CRAD, Management - Oak Ridge National Laboratory High Flux Isotope Reactor  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Oak Ridge National Laboratory High Flux Isotope Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Management portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR More Documents & Publications

10

CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Oak Ridge National Laboratory High Flux Isotope Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Engineering Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory, High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR More Documents & Publications

11

CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Contractor ORR Reactor Contractor ORR CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Maintenance Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Maintenance - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR More Documents & Publications CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor

12

CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Contractor ORR Reactor Contractor ORR CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Training Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Training - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR More Documents & Publications CRAD, Conduct of Operations - Oak Ridge National Laboratory High Flux

13

Neutronics Modeling of the High Flux Isotope Reactor using COMSOL  

Science Conference Proceedings (OSTI)

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

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

2011-01-01T23:59:59.000Z

14

Studies of Past Operations at the High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

During the past year, two topics related to past operations of the High Flux Isotope Reactor (HFIR) were reviewed in response to on-going programs at Oak Ridge National Laboratory (ORNL). Currently, studies are being conducted to determine if HFIR can be converted from high enriched uranium (HEU) fuel to low enriched uranium (LEU). While the basis for conversion is the current performance of the reactor, redesign studies revealed an apparent slight degradation in performance of the reactor over its 40 year lifetime. A second program requiring data from HFIR staff is the Integrated Facility Disposition Project (IFDP). The IFDP is a program that integrates environmental cleanup with modernization and site revitalization plans and projects. Before a path of disposal can be established for discharged HFIR beryllium reflector regions, the reflector components must be classified as to type of waste and specifically, determine if they are transuranic waste.

Chandler, David [ORNL; Primm, Trent [ORNL

2009-01-01T23:59:59.000Z

15

Performance and safety parameters for the high flux isotope reactor  

Science Conference Proceedings (OSTI)

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

Ilas, G. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831-6172 (United States); Primm III, T. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831-6172 (United States); Primm Consulting, LLC, 945 Laurel Hill Road, Knoxville, TN 37923 (United States)

2012-07-01T23:59:59.000Z

16

Performance and Safety Parameters for the High Flux Isotope Reactor  

SciTech Connect

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

Ilas, Germina [ORNL; Primm, Trent [Primm Consulting, LLC

2012-01-01T23:59:59.000Z

17

Calculation of heating values for the high flux isotope reactor  

Science Conference Proceedings (OSTI)

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

Peterson, J.; Ilas, G. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831-6172 (United States)

2012-07-01T23:59:59.000Z

18

Calculation of Heating Values for the High Flux Isotope Reactor  

SciTech Connect

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

Peterson, Joshua L [ORNL; Ilas, Germina [ORNL

2012-01-01T23:59:59.000Z

19

Fabrication of control rods for the High Flux Isotope Reactor  

SciTech Connect

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.

Sease, J.D.

1998-03-01T23:59:59.000Z

20

The HIgh Flux Isotope Reactor: Past, Present, and Future  

Science Conference Proceedings (OSTI)

HFIR construction began in 1965 and completed in 1966. During the first 15 years of operation, the heavy actinide isotope production mission was dominant. HFIR is now positioned as one of the most versataile research reactors in the world.

Beierschmitt, Kelly J [ORNL; Farrar, Mike B [ORNL

2009-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

High Flux Isotope Reactor cold neutron source reference design concept  

SciTech Connect

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.

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

1998-05-01T23:59:59.000Z

22

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

SciTech Connect

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

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

2002-02-26T23:59:59.000Z

23

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

Science Conference Proceedings (OSTI)

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)

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)

2012-07-01T23:59:59.000Z

24

A brief History of Neutron Scattering at the Oak Ridge High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

Neutron scattering at the Oak Ridge National Laboratory dates back to 1945 when Ernest Wollan installed a modified x-ray diffractometer on a beam port of the original graphite reactor. Subsequently, Wollan and Clifford Shull pioneered neutron diffraction and laid the foundation for an active neutron scattering effort that continued through the 1950s, using the Oak Ridge Research reactor after 1958, and, starting in 1966, the High Flux Isotope Reactor, or HFIR.

Nagler, Stephen E [ORNL; Mook Jr, Herbert A [ORNL

2008-01-01T23:59:59.000Z

25

Advanced LWR Fuel Testing Capabilities in the ORNL High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

A new test capability for the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) is being developed that will allow testing of advanced nuclear fuels and cladding materials under prototypic light-water reactor (LWR) operating conditions in less time than it takes in other research reactors. This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiments currently planned to start in late 2008.

Ott, Larry J [ORNL; McDuffee, Joel Lee [ORNL; Spellman, Donald J [ORNL

2008-01-01T23:59:59.000Z

26

Experimental and Computational Study of the Flux Spectrum in Materials Irradiation Facilities of the High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

This report compares the available experimental neutron flux data in the High Flux Isotope Reactor (HFIR) to computational models of the HFIR loosely based on the experimental loading of cycle 400. Over the last several decades, many materials irradiation experiments have included fluence monitors which were subsequently used to reconstruct a coarse-group energy-dependent flux spectrum. Experimental values for thermal and fast neutron flux in the flux trap about the midplane are found to be 1.78 0.27 and 1.05 0:06 1E15 n/cm sec, respectively. The reactor physics code MCNP is used to calculate neutron flux in the HFIR at irradiation locations. The computational results are shown to correspond to closely to experimental data for thermal and fast neutron flux with calculated percent differences ranging from 0:55 13.20%.

McDuffee, Joel Lee [ORNL; Daly, Thomas F [ORNL

2012-01-01T23:59:59.000Z

27

Proposed Program: Reliability-Centered Maintenance (RCM) for the High Flux Isotope Reactor  

E-Print Network (OSTI)

There is a desire to implement a reliability-centered maintenance at the High Flux Isotope Reactor (HFIR) at the Oak-Centered Maintenance (RCM) structure is proposed for implementation at the HFIR. This proposed RCM structure is based on widely used and accepted industry practices. The HFIR primary cleanup system is used to provide specific

28

Homogeneous fast-flux isotope-production reactor  

DOE Patents (OSTI)

A method is described for producing tritium in a liquid metal fast breeder reactor. Lithium target material is dissolved in the liquid metal coolant in order to facilitate the production and removal of tritium.

Cawley, W.E.; Omberg, R.P.

1982-08-19T23:59:59.000Z

29

The ORNL High Flux Isotope Reactor and New Advanced Fuel Testing Capabilities  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy s High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), was originally designed (in the 1960s) primarily as a part of the overall program to produce transuranic isotopes for use in the heavy-element research program of the United States. Today, the reactor is a highly versatile machine, producing medical and transuranic isotopes and performing materials test experimental irradiations and neutron-scattering experiments. The ability to test advanced fuels and cladding materials in a thermal neutron spectrum in the United States is limited, and a fast-spectrum irradiation facility does not currently exist in this country. The HFIR has a distinct advantage for consideration as a fuel/cladding irradiation facility because of the extremely high neutron fluxes that this reactor provides over the full thermal- to fast-neutron energy range. New test capabilities have been developed that will allow testing of advanced nuclear fuels and cladding materials in the HFIR under prototypic light-water reactor (LWR) and fast-reactor (FR) operating conditions.

Ott, Larry J [ORNL; McDuffee, Joel Lee [ORNL

2011-01-01T23:59:59.000Z

30

Review of the Oak Ridge National Laboratory High Flux Isotope Reactor Implementation Verification Review Processes  

NLE Websites -- All DOE Office Websites (Extended Search)

Independent Oversight Review of the Independent Oversight Review of the Oak Ridge National Laboratory High Flux Isotope Reactor Implementation Verification Review Processes May 2011 January 2013 Office of Safety and Emergency Management Evaluations Office of Enforcement and Oversight Office of Health, Safety and Security U. S. Department of Energy Table of Contents 1.0 Purpose ................................................................................................................................................. 1 2.0 Background........................................................................................................................................... 1 3.0 Scope..................................................................................................................................................... 2

31

Review of the Oak Ridge National Laboratory High Flux Isotope Reactor Implementation Verification Review Processes  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Independent Oversight Review of the Independent Oversight Review of the Oak Ridge National Laboratory High Flux Isotope Reactor Implementation Verification Review Processes May 2011 January 2013 Office of Safety and Emergency Management Evaluations Office of Enforcement and Oversight Office of Health, Safety and Security U. S. Department of Energy Table of Contents 1.0 Purpose ................................................................................................................................................. 1 2.0 Background........................................................................................................................................... 1 3.0 Scope..................................................................................................................................................... 2

32

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

DOE Green Energy (OSTI)

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

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

2009-05-03T23:59:59.000Z

33

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

SciTech Connect

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.

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

2009-04-01T23:59:59.000Z

34

External event Probabilistic Risk Assessment for the High Flux Isotope Reactor (HFIR)  

SciTech Connect

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 {times} 10{sup {minus}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.

Flanagan, G.F.; Johnson, D.H.; Buttemer, D.; Perla, H.F.; Chien, S.H. (Oak Ridge National Lab., TN (USA))

1989-01-01T23:59:59.000Z

35

Recent Studies Related to Past Operations at the High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

During the past year, two topics related to past operations of the High Flux Isotope Reactor (HFIR) were reviewed in response to on-going programs at Oak Ridge National Laboratory (ORNL). Currently, studies are being conducted to determine if HFIR can be converted from high enriched uranium (HEU) fuel to low enriched uranium (LEU). While the basis for conversion is the current performance of the reactor, redesign studies revealed an apparent slight degradation in performance of the reactor over its 40 year lifetime. A second program requiring data from HFIR staff is the Integrated Facility Disposition Project (IFDP). The IFDP is a program that integrates environmental cleanup with modernization and site revitalization plans and projects. Before a path of disposal can be established for discharged HFIR beryllium reflector regions, the reflector components must be classified as to type of waste and specifically, determine if they are transuranic waste.

Chandler, David [ORNL; Primm, Trent [ORNL

2009-01-01T23:59:59.000Z

36

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

SciTech Connect

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

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

2012-01-01T23:59:59.000Z

37

Determining Reactor Neutrino Flux  

E-Print Network (OSTI)

Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understanding of the correlation of uncertainties is required for $\\theta_{13}$ experiments. Precise determination of reactor neutrino flux will also improve the sensitivity of the non-proliferation monitoring and future reactor experiments. We will discuss the flux calculation and recent progresses.

Cao, Jun

2011-01-01T23:59:59.000Z

38

Determining Reactor Neutrino Flux  

E-Print Network (OSTI)

Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understanding of the correlation of uncertainties is required for $\\theta_{13}$ experiments. Precise determination of reactor neutrino flux will also improve the sensitivity of the non-proliferation monitoring and future reactor experiments. We will discuss the flux calculation and recent progresses.

Jun Cao

2011-01-12T23:59:59.000Z

39

PRELIMINARY SOLUTION CRITICAL EXPERIMENTS FOR THE HIGH-FLUX ISOTOPE REACTOR  

DOE Green Energy (OSTI)

The design of the High-Flux Isotope Reactor (HFIR) was supported by a series of preliminary experiments performed at the Oak Ridge Critical Experiments Facility in 1960. The experiments yielded results describing directly some of the expected performance characteristics of the reactor and strengthened the calculational methods used in its design. The critical assembly, like the reactor, was of a flux-trap type in which a central 6-in.-dia column of H/sub 2/O was surrounded by an annulus of fissile material and, in turn, by an annular neutron reflector. The fuel region contained a solution of enriched uranyl nitrate in a mixture of H/sub 2/O and D/sub 2/O and the reflector was a composite of two annuli, the inner one of D/sub 2/O surrounded by one of H/sub 2/O. In most experiments the ends of the assembly were reflected by H/sub 2/O. Important results evaluate the absolute thermal-neutron flux to be expected in the design reactor and describe the flux distributions within this type of assembly. It was also observed that the cadmium ratio along the axis of the assembly was about 100, showing that a highly thermal-neutron flux was truly developed in the trap. It was shown that reduction of the hydrogen density in the central water column to about 80% of its normal value increased the reactivity about 6% and that further hydrogen density reduction decreased the reactivity as the effect of the loss of neutron moderation dominated the effect of the increased coupling across the central column. These considerations are of importance to the safety of the reactor. Additional experiments gave values of the usual critical dimensions and explored the effects on both the dimensions and the flux distributions of changing the concentration of the uranyl nitrate solution, of changing the composition of the solvent, and of adding neutron-absorbing materials to the D/ sub 2/O reflector. These changes were made to alter the neutron properties of the fuel solution over a range including those expected in the reactor itself. (auth)

Fox, J.K.; Gilley, L.W.; Magnuson, D.W.

1963-06-12T23:59:59.000Z

40

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

Science Conference Proceedings (OSTI)

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

Ilas, Dan [ORNL

2012-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Preliminary Notice of Violation - High Flux Isotope Reactor, November 18, 2003  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Department of Energy Department of Energy Washington, DC 20585 November 18, 2003 Dr. Jeffrey Wadsworth [ ] UT-Battelle P.O. Box 2008 Oak Ridge, TN 37831-6255 EA 2003-10 Subject: Preliminary Notice of Violation and Proposed Imposition of Civil Penalty $151,250 Dear Dr. Wadsworth: This letter refers to the Department of Energy's Office of Price-Anderson Enforcement (OE) investigation of the facts and circumstances surrounding nuclear safety work control issues at the High Flux Isotope Reactor (HFIR) and the Radiochemical Engineering Development Center (REDC). Our office initiated this investigation in response to a manual reactor shutdown due to a control cylinder maintenance safety deficiency and operation of a radiological [ ] without required containment, as

42

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

SciTech Connect

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

Mo, S. C.

1998-01-14T23:59:59.000Z

43

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

SciTech Connect

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.

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

2010-10-01T23:59:59.000Z

44

SELECTED STUDIES OF PAST OPERATIONS AT THE ORNL HIGH FLUX ISOTOPE REACTOR  

Science Conference Proceedings (OSTI)

In response to on-going programs at Oak Ridge National Laboratory, two topics related to past operations of the High Flux Isotope Reactor (HFIR) are being reviewed and include determining whether HFIR fuel can be converted from high enriched uranium (HEU) to low enriched uranium (LEU) and determining whether HFIR beryllium reflectors are discharged as transuranic (TRU) waste. The LEU conversion and TRU waste studies are being performed in accordance with the Reduced Enrichment for Research and Test Reactors program and the Integrated Facility Disposition Project, respectively. While assessing data/analysis needs for LEU conversion such as the fuel cycle length and power needed to maintain the current level of reactor performance, a reduction of about 8% (~200 MWD) in the end-of-cycle exposure for HFIR fuel was observed over the lifetime of the reactor (43 years). The SCALE 6.0 computational system was used to evaluate discharged beryllium reflectors and it was discovered if the reflectors are procured according to the current HFIR standard, discharged reflectors would not be TRU waste, but the removable reflector (closest to core) would become TRU waste approximately 40 years after discharge. However, beryllium reflectors have been fabricated with a greater uranium content than that stipulated in the standard and these reflectors would be discharged as TRU waste.

Chandler, David [ORNL; Primm, Trent [ORNL

2010-01-01T23:59:59.000Z

45

Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

Ott, Larry J [ORNL; Ellis, Ronald James [ORNL; McDuffee, Joel Lee [ORNL; Spellman, Donald J [ORNL; Bevard, Bruce Balkcom [ORNL

2009-01-01T23:59:59.000Z

46

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

Science Conference Proceedings (OSTI)

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

Primm, Trent [ORNL

2009-01-01T23:59:59.000Z

47

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

Science Conference Proceedings (OSTI)

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.

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

2009-12-01T23:59:59.000Z

48

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

DOE Green Energy (OSTI)

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

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

1998-06-01T23:59:59.000Z

49

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

Science Conference Proceedings (OSTI)

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is a versatile 85 MW{sub th}, pressurized, light water-cooled and -moderated research reactor. The core consists of two fuel elements, an inner fuel element (IFE) and an outer fuel element (OFE), each constructed of involute fuel plates containing high-enriched-uranium (HEU) fuel ({approx}93 wt% {sup 235}U/U) in the form of U{sub 3}O{sub 8} in an Al matrix and encapsulated in Al-6061 clad. An over-moderated flux trap is located in the center of the core, a large beryllium reflector is located on the outside of the core, and two control elements (CE) are located between the fuel and the reflector. The flux trap and reflector house numerous experimental facilities which are used for isotope production, material irradiation, and cold/thermal neutron scattering. Over the past five decades, the US Department of Energy (DOE) and its agencies have been producing radioisotope power systems used by the National Aeronautics and Space Administration (NASA) for unmanned, long-term space exploration missions. Plutonium-238 is used to power Radioisotope Thermoelectric Generators (RTG) because it has a very long half-life (t{sub 1/2} {approx} 89 yr.) and it generates about 0.5 watts/gram when it decays via alpha emission. Due to the recent shortage and uncertainty of future production, the DOE has proposed a plan to the US Congress to produce {sup 238}Pu by irradiating {sup 237}Np as early as in fiscal year 2011. An annual production rate of 1.5 to 2.0 kg of {sup 238}Pu is expected to satisfy these needs and could be produced in existing national nuclear facilities like HFIR and the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Reactors at the Savannah River Site were used in the past for {sup 238}Pu production but were shut down after the last production in 1988. The nation's {sup 237}Np inventory is currently stored at INL. A plan for producing {sup 238}Pu at US research reactor facilities such as the High Flux Isotope Reactor at ORNL has been initiated by the US DOE and NASA for space exploration needs. Two Monte Carlo-based depletion codes, TRITON (ORNL) and VESTA (IRSN), were used to study the {sup 238}Pu production rates with varying target configurations in a typical HFIR fuel cycle. Preliminary studies have shown that approximately 11 grams and within 15 to 17 grams of {sup 238}Pu could be produced in the first irradiation cycle in one small and one large VXF facility, respectively, when irradiating fresh target arrays as those herein described. Important to note is that in this study we discovered that small differences in assumptions could affect the production rates of Pu-238 observed. The exact flux at a specific target location can have a significant impact upon production, so any differences in how the control elements are modeled as a function of exposure, will also cause differences in production rates. In fact, the surface plot of the large VXF target Pu-238 production shown in Figure 3 illustrates that the pins closest to the core can potentially have production rates as high as 3 times those of pins away from the core, thus implying that a cycle-to-cycle rotation of the targets may be well advised. A methodology for generating spatially-dependent, multi-group self-shielded cross sections and flux files with the KENO and CENTRM codes has been created so that standalone ORIGEN-S inputs can be quickly constructed to perform a variety of {sup 238}Pu production scenarios, i.e. combinations of the number of arrays loaded and the number of irradiation cycles. The studies herein shown with VESTA and TRITON/KENO will be used to benchmark the standalone ORIGEN.

Lastres, Oscar [University of Tennessee, Knoxville (UTK); Chandler, David [University of Tennessee, Knoxville (UTK) & Oak Ridge National Laboratory (ORNL); Jarrell, Joshua J [ORNL; Maldonado, G. Ivan [University of Tennessee, Knoxville (UTK)

2011-01-01T23:59:59.000Z

50

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

Science Conference Proceedings (OSTI)

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

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

2012-01-01T23:59:59.000Z

51

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

DOE Green Energy (OSTI)

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

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

2010-01-01T23:59:59.000Z

52

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

DOE Green Energy (OSTI)

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.

Primm, R.T., III

2003-11-01T23:59:59.000Z

53

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

Science Conference Proceedings (OSTI)

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 and traditional and online focus groups with scientists. The latter include SNS, HFIR, and APS users as well as scientists at ORNL, some of whom had not yet used HFIR and/or SNS. These approaches informed development of the second phase, a quantitative online survey. The survey consisted of 16 questions and 7 demographic categorizations, 9 open-ended queries, and 153 pre-coded variables and took an average time of 18 minutes to complete. The survey was sent to 589 SNS/HFIR users, 1,819 NSLS users, and 2,587 APS users. A total of 899 individuals provided responses for this study: 240 from NSLS; 136 from SNS/HFIR; and 523 from APS. The overall response rate was 18%.

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

2011-03-01T23:59:59.000Z

54

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

Science Conference Proceedings (OSTI)

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.

Ilas, Germina [ORNL; Primm, Trent [ORNL

2011-05-01T23:59:59.000Z

55

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

Science Conference Proceedings (OSTI)

An engineering design study that would enable the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium fuel is ongoing at Oak Ridge National Laboratory. The computational models used to search for a low-enriched uranium (LEU) fuel design that would meet the requirements for the conversion study, and the recent results obtained with these models during FY 2009, are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating high-enriched uranium fuel core. These studies indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations.

Ilas, Germina [ORNL; Primm, Trent [ORNL

2009-11-01T23:59:59.000Z

56

The use of PRA (Probabilistic Risk Assessment) in the management of safety issues at the High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

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 {times} 10{sup {minus}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 138% of the internal event initiated contribution and is dominated by wind initiators. The PRA has provided a basis for the management of a wide range of safety and operation issues at the HFIR. 3 refs., 4 figs., 2 tabs.

Flanagan, G.F.

1990-01-01T23:59:59.000Z

57

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

Science Conference Proceedings (OSTI)

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.

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

2010-02-01T23:59:59.000Z

58

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

Science Conference Proceedings (OSTI)

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

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

2009-03-01T23:59:59.000Z

59

Neutronic Analysis of an Advanced Fuel Design Concept for the High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

This study presents the neutronic analysis of an advanced fuel design concept for the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) that could significantly extend the current fuel cycle length under the existing design and safety criteria. A key advantage of the fuel design herein proposed is that it would not require structural changes to the present HFIR core, in other words, maintaining the same rated power and fuel geometry (i.e., fuel plate thickness and coolant channel dimensions). Of particular practical importance, as well, is the fact that the proposed change could be justified within the bounds of the existing nuclear safety basis. The simulations herein reported employed transport theory-based and exposure-dependent eigenvalue characterization to help improve the prediction of key fuel cycle parameters. These parameters were estimated by coupling a benchmarked three-dimensional MCNP5 model of the HFIR core to the depletion code ORIGEN via the MONTEBURNS interface. The design of an advanced HFIR core with an improved fuel loading is an idea that evolved from early studies by R. D. Cheverton, formerly of ORNL. This study contrasts a modified and increased core loading of 12 kg of 235U against the current core loading of 9.4 kg. The simulations performed predict a cycle length of 39 days for the proposed fuel design, which represents a 50% increase in the cycle length in response to a 25% increase in fissile loading, with an average fuel burnup increase of {approx}23%. The results suggest that the excess reactivity can be controlled with the present design and arrangement of control elements throughout the core's life. Also, the new power distribution is comparable or even improved relative to the current power distribution, displaying lower peak to average fission rate densities across the inner fuel element's centerline and bottom cells. In fact, the fission rate density in the outer fuel element also decreased at these key locations for the proposed design. Overall, it is estimated that the advanced core design could increase the availability of the HFIR facility by {approx}50% and generate {approx}33% more neutrons annually, which is expected to yield sizeable savings during the remaining life of HFIR, currently expected to operate through 2014. This study emphasizes the neutronics evaluation of a new fuel design. Although a number of other performance parameters of the proposed design check favorably against the current design, and most of the core design features remain identical to the reference, it is acknowledged that additional evaluations would be required to fully justify the thermal-hydraulic and thermal-mechanical performance of a new fuel design, including checks for cladding corrosion performance as well as for industrial and economic feasibility.

Xoubi, Ned [ORNL; Primm, Trent [ORNL; Maldonado, G. Ivan [University of Tennessee, Knoxville (UTK)

2009-01-01T23:59:59.000Z

60

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

Science Conference Proceedings (OSTI)

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-350-1, 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 49-1 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 time is accurately calculated, i.e. an integral check. Hence, the time dependent heat generation source terms needed for reactor core thermal hydraulic analysis, if derived from this methodology, have been shown to be accurate for highly enriched uranium (HEU) fuel.

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

2010-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

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

Science Conference Proceedings (OSTI)

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.

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

2011-02-01T23:59:59.000Z

62

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

Science Conference Proceedings (OSTI)

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.

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

2010-02-01T23:59:59.000Z

63

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

SciTech Connect

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.

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

2012-03-01T23:59:59.000Z

64

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

SciTech Connect

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, several engineering proof-of-principle tests would be required. The RERTR program is currently conducting a series of generic fuel qualification tests at the Advanced Test Reactor. A review of these tests and a review of the safety basis for the current, HEU fuel cycle led to the identification of a set of HFIR-specific fuel qualification tests. Much additional study is required to formulate a HFIR-specific fuel qualification plan from this set. However, one such test - creating a graded fuel profile across a flat foil - has been initiated with promising results.

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

2007-11-01T23:59:59.000Z

65

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

Science Conference Proceedings (OSTI)

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

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

2010-01-01T23:59:59.000Z

66

CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Nuclear Safety Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor

67

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

QUALITY ASSURANCE (QA) QUALITY ASSURANCE (QA) OBJECTIVE QA-1: The RRD QA program has been appropriately modified to reflect the CS modification and its reactor interface, and sufficient numbers of qualified QA personnel are provided to ensure services are adequate to support reactor operation. The QA functions, assignments, responsibilities, and reporting relationships are clearly defined, understood, and effectively implemented with line management control of safety. QA personnel exhibit awareness of the applicable requirements pertaining to reactor operation with the CS and the associated hazards. Through their actions, they have demonstrated a high-priority commitment to comply with these requirements. The level of knowledge of QA personnel related to reactor

68

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

EMERGENCY PREPAREDNESS (EP) EMERGENCY PREPAREDNESS (EP) OBJECTIVE EP-1: A routine drill program and emergency operations drill program, including program records, have been established and implemented. (Core Requirement 11) Criteria * Reactor operation with the CS has been appropriately incorporated into the emergency preparedness hazards analysis and emergency response procedures. * The implemented routine and emergency operations drill program, including program records, have incorporated the CS SSCs and the CS's operation, hazards, and reactor interface. * Proficiency to appropriately respond to incidents and accidents associated with reactor operation has been demonstrated through the implemented routine and emergency operations drill program. Approach Record Review: Examine ORNL/RRD/INT-114, HFIR Emergency Planning Hazards

69

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

ENGINEERING (ENG) ENGINEERING (ENG) OBJECTIVE ENG-1: The engineering program has been appropriately modified to reflect the CS modification and its reactor interface, sufficient numbers of qualified engineering personnel are provided, and adequate facilities and equipment are available to ensure engineering services are adequate to support reactor and CS operations. The engineering functions, assignments, responsibilities, and reporting relationships are clearly defined, understood, and effectively implemented with line management control of safety. Engineering personnel exhibit awareness of the applicable requirements pertaining to reactor operation with the CS and with CS operations and hazards. Through their actions, they have demonstrated a high-priority commitment

70

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Training & Qualification Training & Qualification OBJECTIVE TR-1: The selection, training and qualification programs associated with CS modifications, operation, hazards, and reactor operations with the hydrogen- moderated CS have been established, documented, and implemented. The selection process and applicable position-specific training for managers and staff, associated with CS modifications and hazards, and reactor operations with the hydrogen- moderated CS ensures competence commensurate with responsibilities (the training and qualification program encompasses the range of duties required to be performed). (CR - 1, CR - 2, CR - 6) Criteria * The Training program is established, documented, and functioning to support reactor operations with the CS modification. Functions, responsibilities, and

71

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

OPERATIONS OPERATIONS OBJECTIVE OP-1: Operations staff and management exhibit awareness of applicable requirements pertaining to CS operation, hazards, and reactor operations with the hydrogen-moderated CS. Through their actions, they have demonstrated a high-priority commitment to comply with these requirements. The level of knowledge of reactor operations and CS system operations managers and staff related to CS operations, hazards, and reactor operations with the hydrogen-moderated CS is adequate based on interviews. Sufficient numbers of qualified reactor operations and CS system operations staff and management are available to conduct and support safe operations with the hydrogen-moderated CS. (CR - 1, CR - 4, CR - 6) Criteria * Minimum staffing requirements have been established for operations and support

72

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

INDUSTRIAL SAFETY AND HYGIENE (IS&H) INDUSTRIAL SAFETY AND HYGIENE (IS&H) OBJECTIVE IS&H-1: The RRD industrial safety and hygiene (IS&H) program has been appropriately modified to reflect the CS modification and its reactor interface, sufficient numbers of qualified IS&H staff and management are provided, and adequate IS&H facilities and equipment are available to ensure services are adequate to support reactor operation with the CS. The IS&H functions, assignments, responsibilities, and reporting relationships are clearly defined, understood, and effectively implemented with line management control of safety. IS&H staff and management exhibit awareness of applicable requirements pertaining to reactor operation with the CS and the associated hazards. Through their actions, they have demonstrated a high-

73

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

Science Conference Proceedings (OSTI)

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

Ilas, Germina [ORNL; Primm, Trent [ORNL

2010-01-01T23:59:59.000Z

74

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

CONFIGURATION MANAGEMENT (CM) CONFIGURATION MANAGEMENT (CM) OBJECTIVE CM-1: The facility systems and procedures, as affected by the facility modifications, are consistent with the description of the facility, procedures, and accident analysis included in the safety basis. (Core Requirement 9) Criteria * The CS and reactor systems affected by the CS and facility modifications are consistent with the description and accident analysis included in the DSA. * The reactor and CS procedures (including system drawings, operating procedures, annunciator response procedures, abnormal operating procedures, emergency operating procedures, surveillance test procedures, and other procedures affected by the CS modification) are consistent with the description and accident analysis included in the DSAs.

75

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

FIRE PROTECTION (FP) FIRE PROTECTION (FP) OBJECTIVE FP-1: The fire protection program has been appropriately modified to reflect the CS and its reactor interface, sufficient numbers of qualified fire protection personnel are available to support operations, and adequate facilities and equipment are available to ensure fire protection services are adequate for operations. The fire protection functions, assignments, responsibilities, and reporting relationships, including those between the line operating organization and the fire protection organization, are clearly defined, understood, and effectively implemented with line management responsibility for control of safety. The level of knowledge of fire protection personnel related to reactor operation with the CS and the associated hazards is adequate.

76

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

ENGINEERING ENGINEERING OBJECTIVE ES-1: The engineering program has been appropriately modified to reflect the CS modification and its reactor interface, sufficient numbers of qualified engineering staff and management are provided, and adequate facilities and equipment are available to ensure services are adequate to conduct and support reactor operations with the hydrogen-moderated CS. Functions, assignments, responsibilities, and reporting relationships are clearly defined, understood, and effectively implemented with line management control of safety. (CR-1, CR-2, CR- 6) Criteria * The engineering organization and associated programs are established and functioning to support the RRD operations organization. Functions, responsibilities, and reporting relationships are clearly defined, understood, and

77

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

MAINTENANCE MAINTENANCE OBJECTIVE MT-1: The maintenance and test programs have been appropriately modified to reflect the CS modification and its reactor interface, sufficient numbers of qualified maintenance and testing staff and management are provided, and adequate facilities and equipment are available to ensure services are adequate to conduct and support reactor operations with the hydrogen-moderated CS. Functions, assignments, responsibilities, and reporting relationships are clearly defined, understood, and effectively implemented with line management control of safety. (CR - 1, CR - 2, CR - 6) Criteria * The maintenance and test programs and organizations are established and functioning to support the RRD operations organization. Functions,

78

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Emergency Preparedness Emergency Preparedness OBJECTIVE EP-1: The emergency preparedness program has been appropriately modified to reflect the CS modification and its reactor interface, sufficient numbers of qualified emergency preparedness staff and management are provided, and adequate facilities and equipment are available to ensure services are adequate to conduct and support reactor operations with the hydrogen-moderated CS. Functions, assignments, responsibilities, and reporting relationships are clearly defined, understood, and effectively implemented with line management control of safety. (CR-1, CR-2, CR-6) Criteria * The emergency preparedness program and organization are established and functioning to support the RRD operations organization. Functions,

79

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

AUTHORIZATION BASIS MANAGEMENT AUTHORIZATION BASIS MANAGEMENT OBJECTIVE AB-1: The nuclear safety program has been appropriately modified to reflect the CS modification and its reactor interface, sufficient numbers of qualified nuclear safety staff and management are provided, and adequate facilities and equipment are available to ensure services are adequate to conduct and support operations with the CS modification. Functions, assignments, responsibilities, and reporting relationships are clearly defined, understood, and effectively implemented with line management control of safety. (CR-1, CR-2, CR-6) Criteria The nuclear safety program and organization are established and functioning to support reactor operations with the CS modification. Functions, responsibilities, and

80

PREPARING THE HIGH FLUX ISOTOPE REACTOR FOR CONVERSION TO LOW ENRICHED URANIUM FUEL ? EXTENDING CYCLE BURNUP  

Science Conference Proceedings (OSTI)

Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting HFIR from high enriched to low enriched uranium (20 wt % 235U) fuel requires extending the end-of-life burnup value for HFIR fuel from the current nominal value of 2200 MWD to 2600 MWD. The current fuel fabrication procedure is discussed and changes that would be required to this procedure are identified. Design and safety related analyses that are required for the certification of a new fuel are identified. Qualification tests and comments regarding the regulatory approval process are provided along with a conceptual schedule.

Primm, Trent [ORNL; Chandler, David [ORNL

2009-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

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

Science Conference Proceedings (OSTI)

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

Chang, S.J.

1997-05-01T23:59:59.000Z

82

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Configuration Management Configuration Management OBJECTIVE CM-1: The CS system and reactor systems affected by the CS modification and associated drawings are consistent with the description and accident analysis included in the DSA and a system to maintain control over their design and modification is established. (CR-9) Criteria * The design requirements have been formally established, documented, and maintained for the CS. * An adequate process has been implemented to ensure that documentation for systems critical to the safety of the facility during operation with the CS exists and is kept current as appropriate for their safety functions, and the documentation is available to the operators. * Cold Source and reactor interface equipment has been included in the

83

Response of aluminum and its alloys to exposure in the high flux isotope reactor  

DOE Green Energy (OSTI)

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 Mg/sub 2/Si precipitate.

Farrell, K.

1983-01-01T23:59:59.000Z

84

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Management and Organization Management and Organization OBJECTIVE MG-1: Line management has integrated within its existing ISM system and implementing mechanisms, programs that appropriately address CS operations, hazards, and reactor interface to assure safe accomplishment of work. Safety management programs of particular interest include the following (CR - 1): * maintenance and testing (addressed by MT-1) * conduct of operations (addressed by OP-1and -5) * training/qualification (addressed by TR-1) * nuclear safety (addressed by AB-2) * emergency management (addressed by EP-1and -2) * configuration management (addressed by ES-3) * fire protection (addressed by ESH-4) * industrial safety and hygiene (addressed by ESH-2) * quality assurance (addressed by ESH-6)

85

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Management Management OBJECTIVE MG-1: Line management has established programs to ensure safe accomplishment of work. Personnel exhibit awareness of public and worker safety, health, and environmental protection requirements, and through their actions, they demonstrate a high-priority commitment to comply with these requirements. (Core Requirements 1 and 2) Criteria * Line management has integrated programs within its existing ISMS and implementing mechanisms that appropriately address the major changes implemented during this outage, notably the CS, in order to continue to assure safe accomplishment of work. * Senior management and RRD management exhibit awareness of the applicable requirements pertaining to reactor operation, with emphasis on the

86

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

RADIOLOGICAL PROTECTION (RP) RADIOLOGICAL PROTECTION (RP) OBJECTIVE RP-1: The RRD radiological protection program has been appropriately modified to reflect the CS modification and its reactor interface, sufficient numbers of qualified radiological protection personnel are provided, and adequate radiological protection facilities and equipment are available to ensure that services are adequate to conduct and support HFIR operation. The radiological protection functions, assignments, responsibilities, and reporting relationships are clearly defined, understood, and effectively implemented with line management control of safety. Radiological protection personnel exhibit awareness of the applicable radiological protection requirements pertaining to HFIR operation and the associated hazards.

87

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

NUCLEAR SAFETY (NS) NUCLEAR SAFETY (NS) OBJECTIVE NS-1: The nuclear safety program has been appropriately modified to reflect the CS modification and its reactor interface, sufficient numbers of qualified nuclear safety personnel are provided, and adequate facilities and equipment are available to ensure that nuclear safety services are adequate to support HFIR operation with the CS. The nuclear safety functions, assignments, responsibilities, and reporting relationships are clearly defined, understood, and effectively implemented with line management control of safety. The level of knowledge of nuclear safety personnel with respect to operation of HFIR with the CS is adequate. (Core Requirements 1, 2, 4, and 6) Criteria * The nuclear safety program is established and functioning to support HFIR

88

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

Science Conference Proceedings (OSTI)

The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are within the bounds of known technology and are adaptable to the high-volume production required to process {approx} 2.5 to 4 tons of U/Mo and produce {approx}16,000 flat plates for U.S. reactors annually ({approx}10,000 of which are needed for HFIR operations). The reference flow sheet is not intended to necessarily represent the best or the most economical way to manufacture a LEU foil fuel for HFIR but simply represents a 'snapshot' in time of technology and is intended to identify the process steps that will likely be required to manufacture a foil fuel. Changes in some of the process steps selected for the reference flow sheet are inevitable; however, no one step or series of steps dominates the overall flow sheet requirements. A result of conceptualizing a reference flow sheet was the identification of the greater number of steps required for a foil process when compared to the dispersion fuel process. Additionally, in most of the foil processing steps, bare uranium must be handled, increasing the complexity of these processing areas relative to current operations. Based on a likely total cost of a few hundred million dollars for a new facility, it is apparent that line item funding will be necessary and could take as much as 8 to 10 years to complete. The infrastructure cost could exceed $100M.

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

2007-09-30T23:59:59.000Z

89

CRAD, Safety Basis - Oak Ridge National Laboratory High Flux Isotope  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Contractor ORR Reactor Contractor ORR CRAD, Safety Basis - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Safety Basis portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Safety Basis - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR More Documents & Publications CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor

90

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

SciTech Connect

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

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

2012-01-01T23:59:59.000Z

91

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

Science Conference Proceedings (OSTI)

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

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

2011-01-01T23:59:59.000Z

92

MEASUREMENT OF THE NEUTRON SPECTRUM OF THE HB-4 COLD SOURCE AT THE HIGH FLUX ISOTOPE REACTOR AT OAK RIDGE NATIONAL LABORATORY  

DOE Green Energy (OSTI)

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

Robertson, Lee [ORNL; Iverson, Erik B [ORNL

2009-01-01T23:59:59.000Z

93

CRAD, DOE Oversight - Oak Ridge National Laboratory High Flux Isotope  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Oak Ridge National Laboratory High Flux Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, DOE Oversight - Oak Ridge National Laboratory High Flux Isotope Reactor A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a DOE independent oversight assessment of the Oak Ridge National Laboratory programs for oversight of its contractors. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, DOE Oversight - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Nuclear Safety - Oak Ridge National Laboratory High Flux Isotope Reactor

94

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

SciTech Connect

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.

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

2006-02-01T23:59:59.000Z

95

Assumptions and criteria for performing a feasibility study of the conversion of the high flux isotope reactor core to use low-enriched uranium fuel  

SciTech Connect

This paper provides a preliminary estimate of the operating power for the High Flux Isotope Reactor when fuelled with low enriched uranium (LEU). Uncertainties in the fuel fabrication and inspection processes are reviewed for the current fuel cycle [highly enriched uranium (HEU)] and the impact of these uncertainties on the proposed LEU fuel cycle operating power is discussed. These studies indicate that for the power distribution presented in a companion paper in these proceedings, the operating power for an LEU cycle would be close to the current operating power. (authors)

Primm Iii, R. T. [Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831-6399 (United States); Ellis, R. J.; Gehin, J. C. [Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831-6172 (United States); Moses, D. L. [Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831-6050 (United States); Binder, J. L. [Oak Ridge National Laboratory, P. O. Box 2008, Oak Ridge, TN 37831-6162 (United States); Xoubi, N. [Univ. of Cincinnati, Rhodes Hall, ML 72, PO Box 210072, Cincinnati, OH 45221-0072 (United States)

2006-07-01T23:59:59.000Z

96

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

Science Conference Proceedings (OSTI)

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

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

2012-01-01T23:59:59.000Z

97

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

SciTech Connect

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

Ellis, R. J.; Gehin, J. C.; Primm Iii, R. T. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)

2006-07-01T23:59:59.000Z

98

High flux reactor  

DOE Patents (OSTI)

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

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

1988-01-01T23:59:59.000Z

99

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

Science Conference Proceedings (OSTI)

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

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

1989-01-01T23:59:59.000Z

100

CRAD, Fire Protection - Oak Ridge National Laboratory High Flux Isotope  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Fire Protection - Oak Ridge National Laboratory High Flux Fire Protection - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Fire Protection - Oak Ridge National Laboratory High Flux Isotope Reactor February 2006 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2006 Commencement of Operations assessment of the Fire Protection program at the Idaho National Laboratory, Idaho Accelerated Retrieval Project Phase II. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Fire Protection - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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101

Use of a cryogenic sampler to measure radioactive gas concentrations in the main off-gas system at a high-flux isotope reactor  

Science Conference Proceedings (OSTI)

A method for measuring gamma-emitting radioactive gases in air has been developed at Oak Ridge National Laboratory (ORNL). This method combines a cryogenic air-sample collector with a high-purity germanium (HPGe) gamma spectroscopy system. This methodology was developed to overcome the inherently difficult collection and detection of radioactive noble gases. The cryogenic air-sampling system and associated HPGe detector has been used to measure the concentration of radioactive gases in the primary coolant main off-gas system at ONRL's High-Flux Isotope Reactor (HFIR). This paper provides: (1) a description of the cryogenic sampler, the radionuclide detection technique, and a discussion of the effectiveness of sampling and detection of gamma-emitting noble gases; (2) a brief description of HFIR and its associated closed high off-gas system; and (3) quantification of gamma-emitting gases present in the off-gas of the HFIR primary core coolant (e.g. radioisotopes of argon, xenon, and krypton).

Berven, B.A.; Perdue, P.T.; Kark, J.B.; Gibson, M.O.

1982-01-01T23:59:59.000Z

102

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

SciTech Connect

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

Not Available

1994-03-01T23:59:59.000Z

103

Isotope correlation studies relative to high enrichment test reactor fuels  

SciTech Connect

Several correlations of fission product isotopic ratios with atom percent fission and neutron flux, for highly enriched /sup 235/U fuel irradiated in two different water moderated thermal reactors, have been evaluated. In general, excellent correlations were indicated for samples irradiated in the same neutron spectrum; however, significant differences in the correlations were noted with the change in neutron spectrum. For highly enriched /sup 235/U fuel, the correlation of the isotopic ratio /sup 143/Nd//sup 145 +146/Nd with atom percent fission has wider applicability than the other fission product isotopic ratio evaluated. The /sup 137/Cs//sup 135/Cs atom ratio shows promise for correlation with neutron flux. Correlations involving heavy element ratios are very sensitive to the neutron spectrum.

Maeck, W.J.; Tromp, R.L.; Duce, F.A.; Emel, W.A.

1978-06-01T23:59:59.000Z

104

Sandia National Laboratories Medical Isotope Reactor concept.  

SciTech Connect

This report describes the Sandia National Laboratories Medical Isotope Reactor and hot cell facility concepts. The reactor proposed is designed to be capable of producing 100% of the U.S. demand for the medical isotope {sup 99}Mo. The concept is novel in that the fuel for the reactor and the targets for the {sup 99}Mo production are the same. There is no driver core required. The fuel pins that are in the reactor core are processed on a 7 to 21 day irradiation cycle. The fuel is low enriched uranium oxide enriched to less than 20% {sup 235}U. The fuel pins are approximately 1 cm in diameter and 30 to 40 cm in height, clad with Zircaloy (zirconium alloy). Approximately 90 to 150 fuel pins are arranged in the core in a water pool {approx}30 ft deep. The reactor power level is 1 to 2 MW. The reactor concept is a simple design that is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days. The fuel fabrication, reactor design and operation, and {sup 99}Mo production processing use well-developed technologies that minimize the technological and licensing risks. There are no impediments that prevent this type of reactor, along with its collocated hot cell facility, from being designed, fabricated, and licensed today.

Coats, Richard Lee; Dahl, James J.; Parma, Edward J., Jr.

2010-04-01T23:59:59.000Z

105

Strategic Isotope Production | ornl.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

Strategic Isotope Production SHARE Strategic Isotope Production ORNL's unique facilities at the High Flux Isotope Reactor (HFIR), Radiochemical Engineering Development Center...

106

Final Report on Isotope Ratio Techniques for Light Water Reactors  

SciTech Connect

The Isotope Ratio Method (IRM) is a technique for estimating the energy or plutonium production in a fission reactor by measuring isotope ratios in non-fuel reactor components. The isotope ratios in these components can then be directly related to the cumulative energy production with standard reactor modeling methods.

Gerlach, David C.; Gesh, Christopher J.; Hurley, David E.; Mitchell, Mark R.; Meriwether, George H.; Reid, Bruce D.

2009-07-01T23:59:59.000Z

107

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

SciTech Connect

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

Hauptman, H.M.; Petro, J.N.; Jacobi, O. [and others

1995-04-01T23:59:59.000Z

108

The Effects of Flux Spectrum Perturbation on Transmutation of Actinides: Optimizing the Production of Transcurium Isotopes  

SciTech Connect

This research presented herein involves the optimization of transcurium production in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Due to the dependence of isotope cross sections on incoming neutron energy, the efficiency with which an isotope is transmuted is highly dependent upon the flux spectrum. There are certain energy bands in which the rate of fission of transcurium production feedstock materials is minimized, relative to the rate of non-fission absorptions. It is proposed that by perturbing the flux spectrum, it is possible to increase the amount of key isotopes, such as 249Bk and 252Cf, that are produced during a transmutation cycle, relative to the consumption of feedstock material. This optimization process is carried out by developing an iterative objective framework involving problem definition, flux spectrum and cross section analysis, simulated transmutation, and analysis of final yields and transmutation parameters. It is shown that it is possible to perturb the local flux spectrum in the transcurium target by perturbing the composition of the target. It is further shown that these perturbations are able to alter the target yields in a non-negligible way. Future work is necessary to develop the optimization framework, and identify the necessary algorithms to update the problem definition based upon progress towards the optimization goals.

Hogle, Susan L [ORNL; Maldonado, G Ivan [ORNL; Alexander, Charles W [ORNL

2012-01-01T23:59:59.000Z

109

High Flux Isotope Reactor | ORNL Neutron Sciences  

NLE Websites -- All DOE Office Websites (Extended Search)

Home Facilities HFIR How to Work with HFIR How to Work with HFIR HFIR Workflow Please contact the experiment interface or coordinator for additional information and...

110

Molecular isotopic effects on coupled electronic and nuclear fluxes  

SciTech Connect

A full quantum treatment shows that coupled electronic and nuclear fluxes exhibit a strong sensitivity to a small mass change in a vibrating molecule. This has been exemplified with the existing isotopes of H{sub 2}{sup +} as well as few fictitious ones. We find that the fluxes undergo a significant change as one goes from one isotope of reduced mass {mu} to another. Other well-defined observables are likewise affected. It turns out that as a general rule, the heavier the isotope, the larger the flux, the smaller the dispersion, and the longer the revival period. While we were able to confirm analytically that the time at the first turning point scales as {radical}({mu}) and that the revival period changes linearly with {mu}, the mechanism of other observables remains subtle as the result of quantum interference highlighted by the pronounced difference observed on the dispersion pattern.

Kenfack, A.; Paulus, B. [Physikalische und Theoretische Chemie, Institut fuer Chemie und Biochemie, Freie Universitaet Berlin, D-14195 Berlin (Germany); Barth, I. [Physikalische und Theoretische Chemie, Institut fuer Chemie und Biochemie, Freie Universitaet Berlin, D-14195 Berlin (Germany); Max-Born-Institut, Max-Born-Strasse 2A, D-12489 Berlin (Germany); Marquardt, F. [Visualisierung und Datenanalyse, Zuse Institut Berlin, D-14195 Berlin (Germany); Fachbereich Mathematik, Freie Universitaet Berlin, D-14195 Berlin (Germany)

2010-12-15T23:59:59.000Z

111

Small-Scale Reactor for the Production of Medical Isotopes  

Small-Scale Reactor for the Production of Medical Isotopes IP Home; Search/Browse Technology ... Drawing upon proven technology with minimal research effort required;

112

High Flux Beam Reactor | Environmental Restoration Projects | BNL  

NLE Websites -- All DOE Office Websites (Extended Search)

Environmental Protection Division Environmental Protection Division Home Reactor Projects Celebrating DOE's Cleanup Accomplishments (PDF) Brookhaven Graphite Research Reactor(BGRR) BGRR Overview BGRR Complex Description Decommissioning Decision BGRR Complex Cleanup Actions BGRR Documents BGRR Science & Accomplishments High Flux Beam Reactor (HFBR) HFBR Overview HFBR Complex Description Decommissioning Decision HFBR Complex Cleanup Actions HFBR Documents HFBR Science & Accomplishments Groundwater Protection Group Environmental Protection Division Contact > See also: HFBR Science & Accomplishments High Flux Beam Reactor Under the U.S. Department of Energy (DOE), the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) underwent stabilization and partial decommissioning to prepare the HFBR confinement for long-term safe

113

Research and Medical Isotope Reactor Supply | Y-12 National Security...  

NLE Websites -- All DOE Office Websites (Extended Search)

Research and Medical ... Research and Medical Isotope Reactor Supply Our goal is to fuel research and test reactors with low-enriched uranium. Y-12 tops the short list of the...

114

High Flux Beam Reactor | Environmental Restoration Projects | BNL  

NLE Websites -- All DOE Office Websites (Extended Search)

Why is the High Flux Beam Reactor Being Decommissioned? Why is the High Flux Beam Reactor Being Decommissioned? HFBR The High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) is being decommissioned because the Department of Energy (DOE) decided in 1999 that it would be permanently closed. The reactor was shut down in 1997 after tritium from a leak in the spent-fuel pool was found in the groundwater. The HFBR, which had operated from 1965 to 1996, was used solely for scientific research, providing neutrons for materials science, chemistry, biology, and physics experiments. The reactor was shut down for routine maintenance in November of 1996. In January 1997, tritium, a radioactive form of hydrogen and a by-product of reactor operations, was found in groundwater monitoring wells immediately south of the HFBR. The tritium

115

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

Science Conference Proceedings (OSTI)

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

NONE

1995-06-01T23:59:59.000Z

116

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

Science Conference Proceedings (OSTI)

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

NONE

1995-07-01T23:59:59.000Z

117

Spheromak reactor with poloidal flux-amplifying transformer  

DOE Patents (OSTI)

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.

Furth, Harold P. (Princeton, NJ); Janos, Alan C. (East Windsor, NJ); Uyama, Tadao (Osaka, JP); Yamada, Masaaki (Lawrenceville, NJ)

1987-01-01T23:59:59.000Z

118

Spheromak reactor with poloidal flux-amplifying transformer  

DOE Patents (OSTI)

It is an object of the present invention to provide for improved generation and sustainment of an energetic plasma in a spheromak fusion reactor. A large poloidal magnetic flux is inductively induced in a spheromak-shaped plasma utilizing a reduced magnetic field-generating current in a current-carrying flux core.

Furth, H.P.; Janos, A.C.; Uyama, T.; Yamada, M.

1986-05-21T23:59:59.000Z

119

Simulation of the SONGS Reactor Antineutrino Flux Using DRAGON  

E-Print Network (OSTI)

For reactor antineutrino experiments, a thorough understanding of the fuel composition and isotopic evolution is of paramount importance for the extraction of $\\theta_{13}$. To accomplish these goals, we employ the deterministic lattice code DRAGON, and analyze the instantaneous antineutrino rate from the San Onofre Nuclear Generating Station (SONGS) Unit 2 reactor in California. DRAGON's ability to predict the rate for two consecutive fuel cycles is examined.

Jones, C L

2011-01-01T23:59:59.000Z

120

Scoping assessment on medical isotope production at the Fast Flux Test Facility  

Science Conference Proceedings (OSTI)

The Scoping Assessment addresses the need for medical isotope production and the capability of the Fast Flux Test Facility to provide such isotopes. Included in the discussion are types of isotopes used in radiopharmaceuticals, which types of cancers are targets, and in what way isotopes provide treatment and/or pain relief for patients.

Scott, S.W.

1997-08-29T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Neutron flux profile monitor for use in a fission reactor  

DOE Patents (OSTI)

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

Kopp, Manfred K. (Oak Ridge, TN); Valentine, Kenneth H. (Lenoir City, TN)

1983-01-01T23:59:59.000Z

122

Research Reactors Division | ORNL Neutron Sciences  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactors Division (RRD) is responsible for operation of the High Flux Isotope Reactor (HFIR). Operating at 85 MW, HFIR is the highest flux reactor-based source of neutrons for...

123

Nuclear reactor fissile isotopes antineutrino spectra  

E-Print Network (OSTI)

Positron spectrum from inverse beta decay reaction on proton was measured in 1988-1990 as a result of neutrino exploration experiment. The measured spectrum has the largest statistics and lowest energy threshold between other neutrino experiments made that time at nuclear reactors. On base of the positron spectrum the standard antineutrino spectrum for typical reactor fuel composition was restored. In presented analysis the partial spectra forming this standard spectrum were extracted using specific method. They could be used for neutrino experiments data analysis made at any fuel composition of reactor core.

Sinev, V

2012-01-01T23:59:59.000Z

124

Nuclear reactor fissile isotopes antineutrino spectra  

E-Print Network (OSTI)

Positron spectrum from inverse beta decay reaction on proton was measured in 1988-1990 as a result of neutrino exploration experiment. The measured spectrum has the largest statistics and lowest energy threshold between other neutrino experiments made that time at nuclear reactors. On base of the positron spectrum the standard antineutrino spectrum for typical reactor fuel composition was restored. In presented analysis the partial spectra forming this standard spectrum were extracted using specific method. They could be used for neutrino experiments data analysis made at any fuel composition of reactor core.

V. Sinev

2012-07-30T23:59:59.000Z

125

Decommissioning of the high flux beam reactor at Brookhaven Lab  

Science Conference Proceedings (OSTI)

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

Hu, J.P. [National Synchrotron Light Source, Brookhaven Laboratory, Upton, NY 11973 (United States); Reciniello, R.N. [Radiological Control Div., Brookhaven Laboratory, Upton, NY 11973 (United States); Holden, N.E. [National Nuclear Data Center, Brookhaven Laboratory, Upton, NY 11973 (United States)

2011-07-01T23:59:59.000Z

126

INEEL/EXT-01-01623 MODULAR PEBBLE-BED REACTOR PROJECT  

E-Print Network (OSTI)

in the early 1990s. Fuel compacts were irradiated at the High Flux Isotope Reactor (HFIR) and the Advanced Test

127

Thermal-hydraulic simulation of natural convection decay heat removal in the High Flux Isotope Reactor (HFIR) using RELAP5 and TEMPEST: Part 2, Interpretation and validation of results  

SciTech Connect

The RELAP5/MOD2 code was used to predict the thermal-hydraulic behavior of the HFIR core during decay heat removal through boiling natural circulation. The low system pressure and low mass flux values associated with boiling natural circulation are far from conditions for which RELAP5 is well exercised. Therefore, some simple hand calculations are used herein to establish the physics of the results. The interpretation and validation effort is divided between the time average flow conditions and the time varying flow conditions. The time average flow conditions are evaluated using a lumped parameter model and heat balance. The Martinelli-Nelson correlations are used to model the two-phase pressure drop and void fraction vs flow quality relationship within the core region. Systems of parallel channels are susceptible to both density wave oscillations and pressure drop oscillations. Periodic variations in the mass flux and exit flow quality of individual core channels are predicted by RELAP5. These oscillations are consistent with those observed experimentally and are of the density wave type. The impact of the time varying flow properties on local wall superheat is bounded herein. The conditions necessary for Ledinegg flow excursions are identified. These conditions do not fall within the envelope of decay heat levels relevant to HFIR in boiling natural circulation. 14 refs., 5 figs., 1 tab.

Ruggles, A.E.; Morris, D.G.

1989-01-01T23:59:59.000Z

128

Determining Reactor Flux from Xenon-136 and Cesium-135 in Spent Fuel  

E-Print Network (OSTI)

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

Hayes, A C

2012-01-01T23:59:59.000Z

129

Determining Reactor Flux from Xenon-136 and Cesium-135 in Spent Fuel  

E-Print Network (OSTI)

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

A. C. Hayes; Gerard Jungman

2012-05-30T23:59:59.000Z

130

Rebuilding the Brookhaven high flux beam reactor: A feasibility study  

SciTech Connect

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.

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

1995-01-01T23:59:59.000Z

131

Research Reactors Division | Neutron Science | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

is responsible for operation of the High Flux Isotope Reactor. Operating at 85 MW, HFIR is the highest flux reactor-based source of neutrons for research in the United States,...

132

Research reactors - an overview  

SciTech Connect

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

West, C.D.

1997-03-01T23:59:59.000Z

133

Design and optimization of a high thermal flux research reactor via Kriging-based algorithm  

E-Print Network (OSTI)

In response to increasing demands for the services of research reactors, a 5 MW LEU-fueled research reactor core is developed and optimized to provide high thermal flux within specified limits upon thermal hydraulic ...

Kempf, Stephanie Anne

2011-01-01T23:59:59.000Z

134

Fast Flux Test Reactor: Re-evaluation of the Department's Approach...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Sites Power Marketing Administration Other Agencies You are here Home Fast Flux Test Reactor: Re-evaluation of the Department's Approach to Deactivation, Decontamination,...

135

Hydrogen Isotope Separation System for the Tokamak Experimental Power Reactor  

SciTech Connect

An isotopic separation system for processing the fuel in the Tokamak Experimental Power Reactor is described. Two cryogenic distillation columns are used in sequence to recover 80% of the hydrogen from a fuel mixture originally containing equal parts of deuterium and tritium with a 1% hydrogen impurity. The hydrogen thus removed contains less than 1/2% tritium, which may be recovered in a separate system designed for that purpose. It is assumed that separation of the deuterium and the tritium is not required. A total tritium inventory of approximately 38,000 Ci (3.8 g) is projected.

Wilkes, W. R.

1976-03-01T23:59:59.000Z

136

Independent Verification of Research Reactor Operation (Analysis of the Georgian IRT-M Reactor by the Isotope Ratio Method)  

SciTech Connect

The U.S. Department of Energy’s Office of Nonproliferation and International Security (NA-24) develops technologies to aid in implementing international nuclear safeguards. The Isotope Ratio Method (IRM) was successfully developed in 2005 – 2007 by Pacific Northwest National Laboratory (PNNL) and the Republic of Georgia’s Andronikashvili Institute of Physics as a generic technology to verify the declared operation of water-moderated research reactors, independent of spent fuel inventory. IRM estimates the energy produced over the operating lifetime of a fission reactor by measuring the ratios of the isotopes of trace impurity elements in non-fuel reactor components.The Isotope Ratio Method is a technique for estimating the energy produced over the operating lifetime of a fission reactor by measuring the ratios of the isotopes of impurity elements in non-fuel reactor components.

Cliff, John B.; Frank, Douglas P.; Gerlach, David C.; Gesh, Christopher J.; Little, Winston W.; Reid, Bruce D.; Tsiklauri, Georgi V.; Abramidze, Sh; Rostomashvili, Z.; Kiknadze, G.; Dzhavakhishvily, O.; Nabakhtiani, G.

2010-08-11T23:59:59.000Z

137

Preliminary Notice of Violation - High Flux Isotope Reactor,...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

work processes involving HFIR maintenance planning, implementation and post maintenance testing; (2) work processes involving REDC operations; (3) quality improvement and...

138

Secondary Ionization Mass Spectrometric Analysis of Impurity Element Isotope Ratios in Nuclear Reactor Materials  

Science Conference Proceedings (OSTI)

Secondary ion mass spectrometry (SIMS) analysis has been used to measure isotope ratios of selected impurity elements in irradiated reactor materials. Samples of reactor materials such as graphite or aluminum alloys are obtained from fuel channels or supporting materials. During reactor operations and fuel burn up, some isotopic abundances change due to nuclear reactions and provide sensitive indicators of neutron fluence. The rate of change is related to cross section for a particular isotope. Different isotopes can be used as indicators of burn up during different stages in the reactor operating history. Isotope ratios of B are useful indicators for low burnup stages early in reactor operations, Ti isotope ratios are useful at later burn up stages, and Cl isotope ratios are useful in both early and later stages. Knowledge of the sample position within the reactor also yields information on the fluence shape or profile. In a sequence of samples from one reactor, 10B/11B ratios decreased from near natural values of 0.25 to blasting, plasma etching, and vacuum furnace treatment.

Gerlach, David C.; Cliff, John B.; Hurley, David E.; Reid, Bruce D.; Little, Winston W.; Meriwether, George H.; Wickham, Anthony J.; Simmons, Tere A.

2006-07-30T23:59:59.000Z

139

Level 1 Tornado PRA for the High Flux Beam Reactor  

Science Conference Proceedings (OSTI)

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}.

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

1994-05-01T23:59:59.000Z

140

POLOIDAL FLUX LINKAGE REQUIREMENTS FOR THE INTERNATIONAL THERMONUCLEAREXPERIMENTAL REACTOR  

E-Print Network (OSTI)

the poloidal flux linkage requirements for the current ramp-up and for the flat-top phase of the proposed [2] and JSOLVER [3] to calculate the flux linkage requirements during the current ramp-up and steady regarding the plasma ramp-up time, and the amount of flux linkage change that the poloidal field coil system

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Prediction of the reactor antineutrino flux for the Double Chooz experiment  

E-Print Network (OSTI)

This thesis benchmarks the deterministic lattice code, DRAGON, against data, and then applies this code to make a prediction for the antineutrino flux from the Chooz BI and B2 reactors. Data from the destructive assay of ...

Jones, Christopher LaDon

2012-01-01T23:59:59.000Z

142

Neutron Flux Measurements and Calculations in the Gamma Irradiation Facility Using MCNPX.  

E-Print Network (OSTI)

??The gamma irradiation facility at the High Flux Isotope Reactor (HFIR)is used to deliver a pure gamma dose to any target of interest. in addition… (more)

Giuliano, Dominic Richard

2010-01-01T23:59:59.000Z

143

CRAD, Radiological Controls - Oak Ridge National Laboratory High Flux  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

High High Flux Isotope Reactor CRAD, Radiological Controls - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Radiation Protection Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Radiological Controls - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Engineering - Oak Ridge National Laboratory High Flux Isotope Reactor

144

PRELIMINARY HAZARDS SUMMARY REPORT ON THE BROOKHAVEN HIGH FLUX BEAM RESEARCH REACTOR  

SciTech Connect

The High Flux Beam Reactor, HFBR, is cooled, moderated, and reflected by heavy water and designed to produce 40 Mw with a total epithermal flux of ~1.6 X 10/sup 15/cm/sup -2/ sec/sup -1/ and a flector thermal maximum flux of 7 X 10/sup 14/ cm/sup -2/ sec/sup -1/, using a core formed by ETR plate-type fuel elements in a close-packed array. The hazards summary is given in terms of site description, reactor design, building design, plant operation, disposal of radioactive wastes and effluents, and safety analysis. (B.O.G.)

Hendrie, J.M.; Kouts, H.J.C.

1961-05-01T23:59:59.000Z

145

Nested reactor chamber and operation for Hg-196 isotope separation process  

DOE Patents (OSTI)

The present invention is directed to an apparatus for use in .sup.196 Hg separation and its method of operation. Specifically, the present invention is directed to a nested reactor chamber useful for .sup.196 Hg isotope separation reactions avoiding the photon starved condition commonly encountered in coaxial reactor systems.

Grossman, Mark W. (Belmont, MA)

1991-01-01T23:59:59.000Z

146

Measuring of fissile isotopes partial antineutrino spectra in direct experiment at nuclear reactor  

E-Print Network (OSTI)

The direct measuring method is considered to get nuclear reactor antineutrino spectrum. We suppose to isolate partial spectra of the fissile isotopes by using the method of antineutrino spectrum extraction from the inverse beta decay positron spectrum applied at Rovno experiment. This admits to increase the accuracy of partial antineutrino spectra forming the total nuclear reactor spectrum. It is important for the analysis of the reactor core fuel composition and could be applied for non-proliferation purposes.

V. V. Sinev

2009-02-22T23:59:59.000Z

147

Measuring of fissile isotopes partial antineutrino spectra in direct experiment at nuclear reactor  

E-Print Network (OSTI)

The direct measuring method is considered to get nuclear reactor antineutrino spectrum. We suppose to isolate partial spectra of the fissile isotopes by using the method of antineutrino spectrum extraction from the inverse beta decay positron spectrum applied at Rovno experiment. This admits to increase the accuracy of partial antineutrino spectra forming the total nuclear reactor spectrum. It is important for the analysis of the reactor core fuel composition and could be applied for non-proliferation purposes.

Sinev, V V

2009-01-01T23:59:59.000Z

148

Small-Scale Reactor for the Production of Medical Isotopes ...  

Currently, there is a severe worldwide shortage of medical isotopes-specifically Molybdenum 99 (Mo-99) which is essential in cancer treatment, ...

149

High Flux Beam Reactor | Environmental Restoration Projects | BNL  

NLE Websites -- All DOE Office Websites (Extended Search)

Complex Description Complex Description Current HFBR Complex The HFBR complex consists of multiple structures and systems that were necessary to operate and maintain the reactor. The most recognizable features of the complex are the domed reactor confinement building and the distinctive red-and-white stack. Portions of the complex building structures, systems, and components, some of which are underground, were contaminated with radionuclides and chemicals as a result of previous HFBR and Brookhaven Graphite Research Reactor (BGRR) operations. A number of decommissioning and preparation for long-term safe storage actions have been taken including the removal of contaminated structures, hazardous materials, and contaminated equipment and components. The structures and systems, both current and former, are

150

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

SciTech Connect

5098-SR-03-0 FINAL REPORT- INDEPENDENT VERIFICATION SURVEY OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT OUTSIDE AREAS, BROOKHAVEN NATIONAL LABORATORY

P.C. Weaver

2010-12-15T23:59:59.000Z

151

LETTER REPORT INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT FAN HOUSE, BUILDING 704 BNL  

Science Conference Proceedings (OSTI)

5098-LR-01-0 -LETTER REPORT INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT FAN HOUSE, BUILDING 704 BROOKHAVEN NATIONAL LABORATORY

P.C. Weaver

2010-10-22T23:59:59.000Z

152

Neutron-flux profile monitor for use in a fission reactor  

DOE Patents (OSTI)

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

Kopp, M.K.; Valentine, K.H.

1981-09-15T23:59:59.000Z

153

Packed bed reactor for photochemical .sup.196 Hg isotope separation  

DOE Patents (OSTI)

Straight tubes and randomly oriented pieces of tubing having been employed in a photochemical mercury enrichment reactor and have been found to improve the enrichment factor (E) and utilization (U) compared to a non-packed reactor. One preferred embodiment of this system uses a moving bed (via gravity) for random packing.

Grossman, Mark W. (Belmont, MA); Speer, Richard (Reading, MA)

1992-01-01T23:59:59.000Z

154

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

SciTech Connect

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

Pond, R. B.

1998-01-16T23:59:59.000Z

155

CRAD, Emergency Management - Oak Ridge National Laboratory High Flux  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Oak Ridge National Laboratory High Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Emergency Management Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Emergency Management - Oak Ridge National Laboratory High Flux

156

CRAD, Conduct of Operations - Oak Ridge National Laboratory High Flux  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Reactor CRAD, Conduct of Operations - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February, 2007 assessment of the Conduct of Operations Program in preparation for restart of the Oak Ridge National Laboratory, High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Conduct of Operations - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications CRAD, Fire Protection - Oak Ridge National Laboratory High Flux Isotope

157

Evolution of isotopic composition of reprocessed uranium during the multiple recycling in light water reactors with natural uranium feed  

Science Conference Proceedings (OSTI)

A complex approach based on the consistent modeling of neutron-physics processes and processes of cascade separation of isotopes is applied for analyzing physical problems of the multiple usage of reprocessed uranium in the fuel cycle of light water reactors. A number of scenarios of multiple recycling of reprocessed uranium in light water reactors are considered. In the process, an excess absorption of neutrons by the {sup 236}U isotope is compensated by re-enrichment in the {sup 235}U isotope. Specific consumptions of natural uranium for re-enrichment of the reprocessed uranium depending on the content of the {sup 232}U isotope are obtained.

Smirnov, A. Yu., E-mail: a.y.smirnoff@rambler.ru; Sulaberidze, G. A. [National Research Nuclear University MEPhI (Russian Federation); Alekseev, P. N.; Dudnikov, A. A.; Nevinitsa, V. A., E-mail: neva@dhtp.kiae.ru; Proselkov, V. N.; Chibinyaev, A. V. [Russian Research Centre Kurchatov Institute (Russian Federation)

2012-12-15T23:59:59.000Z

158

CRAD, Configuration Management - Oak Ridge National Laboratory High Flux  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Configuration Management - Oak Ridge National Laboratory High Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Configuration Management Program in preparation for restart of the Oak Ridge National Laboratory, High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications

159

CRAD, Environmental Protection - Oak Ridge National Laboratory High Flux  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Environmental Protection - Oak Ridge National Laboratory High Environmental Protection - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Environmental Protection - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Environmental Compliance Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Environmental Protection - Oak Ridge National Laboratory High Flux Isotope Reactor More Documents & Publications

160

CRAD, Emergency Management - Oak Ridge National Laboratory High Flux  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Oak Ridge National Laboratory High Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Emergency Management Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 2 DF WASTE LINE REMOVAL, BNL  

SciTech Connect

5098-SR-02-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 2 DF WASTE LINE REMOVAL, BROOKHAVEN NATIONAL LABORATORY

P.C. Weaver

2010-07-09T23:59:59.000Z

162

Mixing rules for and effects of other hydrogen isotopes and of isotopic swamping on tritium recovery and loss to biosphere from fusion reactors  

DOE Green Energy (OSTI)

Efficient recovery of bred and unburnt tritium from fusion reactors, and control of its migration within reactors and of its escape into the biosphere are essential for self-sufficient fuel cycles and for public, plant personnel, and environmental protection. Tritium in fusion reactors will be mixed with unburnt deuterium and protium introduced by (n,p) reactions and diffusion into coolant loops from steam cycles. Rational design for tritium recovery and escape prevention must acknowledge this fact. Consequences of isotopic admixture are explored, mixing rules for projected fusion reactor dilute-solution conditions are developed, and a rule of thumb regarding their effects on tritium recovery methods is formulated.

Pendergrass, J.H.

1978-01-01T23:59:59.000Z

163

ISOTOPES  

E-Print Network (OSTI)

A Guidebook to Nuclear Reactors, University of Californiaa thermal position of a nuclear reactor followed by analysisproduced by six large nuclear reactors. The power usage per

Lederer, C. Michael

2013-01-01T23:59:59.000Z

164

Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression  

DOE Patents (OSTI)

The invention is a laser or particle-beam-driven fusion reactor system which takes maximum advantage of both the very short pulsed nature of the energy release of inertial confinement fusion (ICF) and the very small volumes within which the thermonuclear burn takes place. The pulsed nature of ICF permits dynamic direct energy conversion schemes such as magnetohydrodynamic (MHD) generation and magnetic flux compression; the small volumes permit very compact blanket geometries. By fully exploiting these characteristics of ICF, it is possible to design a fusion reactor with exceptionally high power density, high net electric efficiency, and low neutron-induced radioactivity. The invention includes a compact blanket design and method and apparatus for obtaining energy utilizing the compact blanket.

Lasche, G.P.

1983-09-29T23:59:59.000Z

165

ISOTOPES  

E-Print Network (OSTI)

uranium, heavy-water-moderated CANDU reactor, as contrastedis important, and in the CANDU power reactor, which uses

Lederer, C. Michael

2013-01-01T23:59:59.000Z

166

Solid tags for identifying failed reactor components  

DOE Patents (OSTI)

A solid tag material which generates stable detectable, identifiable, and measurable isotopic gases on exposure to a neutron flux to be placed in a nuclear reactor component, particularly a fuel element, in order to identify the reactor component in event of its failure. Several tag materials consisting of salts which generate a multiplicity of gaseous isotopes in predetermined ratios are used to identify different reactor components.

Bunch, Wilbur L. (Richland, WA); Schenter, Robert E. (Richland, WA)

1987-01-01T23:59:59.000Z

167

Study of the Neutron Flux and Dpa Attenuation in the Reactor Pressure-Vessel Wall  

Science Conference Proceedings (OSTI)

The study of the neutron flux and dpa attenuation in the reactor pressure vessel (PV) wall presented in this work was performed with state-of-the art methods currently used to determine PV fluxes, the BUGLE-96 cross-section library, and the iron displacement cross sections derived from ENDF/B-VI data. The calculations showed that the RG 1.99, Rev. 2, extrapolation formula predicts slower--and therefore conservative--attenuation of the neutron flux (E > 1MeV) in the PV wall. More importantly, the calculations gave slower attenuation of the dpa rate in the PV wall than the attenuation predicted by the formula. The slower dpa rate attenuation was observed for all the cases considered, which included two different PWRs, and several configurations obtained by varying the PV wall thickness and thermal shield thickness. For example, for a PV wall thickness of {approximately}24 cm, the calculated ratio of the dpa rate at 1/4 and 3/4 of the PV wall thickness to the dpa value on the inner PV surface is {approximately}14% and 19% higher, respectively, than predicted by the RG 1.99, Rev. 2, formula.

Remec, I.

1999-06-01T23:59:59.000Z

168

CRAD, Conduct of Operations - Oak Ridge National Laboratory High Flux  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Contractor ORR Reactor Contractor ORR CRAD, Conduct of Operations - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February, 2007 assessment of the Conduct of Operations Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory, High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Conduct of Operations - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR More Documents & Publications

169

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

Science Conference Proceedings (OSTI)

5098-SR-04-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 5, BROOKHAVEN NATIONAL LABORATORY

P.C. Weaver

2010-11-03T23:59:59.000Z

170

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

SciTech Connect

5098-SR-05-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 1 BROOKHAVEN NATIONAL LABORATORY

E.M. Harpenau

2010-12-15T23:59:59.000Z

171

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

E-Print Network (OSTI)

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

Lawrence, C B

1982-01-01T23:59:59.000Z

172

ISOTOPES  

E-Print Network (OSTI)

Theory of Isotope Separation as Applied to the Large~scale Production of 235 u National Nuclear Energy

Lederer, C. Michael

2013-01-01T23:59:59.000Z

173

Determination of Light Water Reactor Fuel Burnup with the Isotope Ratio Method  

Science Conference Proceedings (OSTI)

For the current project to demonstrate that isotope ratio measurements can be extended to zirconium alloys used in LWR fuel assemblies we report new analyses on irradiated samples obtained from a reactor. Zirconium alloys are used for structural elements of fuel assemblies and for the fuel element cladding. This report covers new measurements done on irradiated and unirradiated zirconium alloys, Unirradiated zircaloy samples serve as reference samples and indicate starting values or natural values for the Ti isotope ratio measured. New measurements of irradiated samples include results for 3 samples provided by AREVA. New results indicate: 1. Titanium isotope ratios were measured again in unirradiated samples to obtain reference or starting values at the same time irradiated samples were analyzed. In particular, 49Ti/48Ti ratios were indistinguishably close to values determined several months earlier and to expected natural values. 2. 49Ti/48Ti ratios were measured in 3 irradiated samples thus far, and demonstrate marked departures from natural or initial ratios, well beyond analytical uncertainty, and the ratios vary with reported fluence values. The irradiated samples appear to have significant surface contamination or radiation damage which required more time for SIMS analyses. 3. Other activated impurity elements still limit the sample size for SIMS analysis of irradiated samples. The sub-samples chosen for SIMS analysis, although smaller than optimal, were still analyzed successfully without violating the conditions of the applicable Radiological Work Permit

Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

2007-11-01T23:59:59.000Z

174

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

Science Conference Proceedings (OSTI)

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

Dautel, W.A.

1996-10-01T23:59:59.000Z

175

CRAD, Configuration Management - Oak Ridge National Laboratory High Flux  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Oak Ridge National Laboratory High Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007, A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Configuration Managment Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory, High Flux Isotope Reactor. CRADs provide a recommended approach and the types of information to gather to assess elements of a DOE contractor's programs. CRAD, Configuration Management - Oak Ridge National Laboratory High Flux

176

A Heterogeneous Sodium Fast Reactor Designed to Transmute Minor Actinide Actinide Waste Isotopes into Plutonium Fuel  

Science Conference Proceedings (OSTI)

An axial heterogeneous sodium fast reactor design is developed for converting minor actinide waste isotopes into plutonium fuel. The reactor design incorporates zirconium hydride moderating rods in an axial blanket above the active core. The blanket design traps the active core’s axial leakage for the purpose of transmuting Am-241 into Pu-238. This Pu-238 is then co-recycled with the spent driver fuel to make new driver fuel. Because Pu-238 is significantly more fissile than Am-241 in a fast neutron spectrum, the fissile worth of the initial minor actinide material is upgraded by its preconditioning via transmutation in the axial targets. Because, the Am-241 neutron capture worth is significantly stronger in a moderated epithermal spectrum than the fast spectrum, the axial targets serve as a neutron trap which recovers the axial leakage lost by the active core. The sodium fast reactor proposed by this work is designed as an overall transuranic burner. Therefore, a low transuranic conversion ratio is achieved by a degree of core flattening which increases axial leakage. Unlike a traditional “pancake” design, neutron leakage is recovered by the axial target/blanket system. This heterogeneous core design is constrained to have sodium void and Doppler reactivity worth similar to that of an equivalent homogeneous design. Because minor actinides are irradiated only once in the axial target region; elemental partitioning is not required. This fact enables the use of metal targets with electrochemical reprocessing. Therefore, the irradiation environment of both drivers and targets was constrained to ensure applicability of the established experience database for metal alloy sodium fast reactor fuels.

Samuel E. Bays

2011-02-01T23:59:59.000Z

177

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

Wigner, E.P.

1958-04-22T23:59:59.000Z

178

Flux  

NLE Websites -- All DOE Office Websites (Extended Search)

5000 5000 6000 7000 8000 Wavelength (Angstroms) Flux (in arbitrary units) SN 1990N SN 1989B SN 1993O SN 1981B SN 1994D SN 1997ap Iron Peak Blends Ca II Si II & Co II Fe II & III Day -7 Day -5 Day -4 Day -2 ± 2 Day 0 Day +2 * -50 0 50 100 150 Observed days from peak Observed I magnitude 27 26 25 24 23 Observed R magnitude 27 26 25 24 Observed I magnitude 27 26 25 24 23 R band Ground-based I band HST I band (b) (c) (a) Pre-SN observation 3.5 4.0 4.5 5.0 5.5 log(cz) 14 16 18 20 22 24 26 effective m B 0.02 0.05 0.1 0.2 0.5 1.0 redshift z Hamuy et al (A.J. 1996) Supernova Cosmology Project 6 8 % 9 0 % 0.5 1.0 1.5 2.0 2.5 3.0 ! M Age < 9.6 Gyr (H = 50 km s -1 Mpc -1 ) No Big Bang 0.0 0.5 1.0 1.5 2.0 2.5 3.0 -3 -2 -1 0 1 2 3 -3 -2 -1 0 1 2 3 ! " z ~ 0 . 4 z = 0 . 8 3 6 8 % 9 0 % 0.5 1.0 1.5 2.0 2.5 3.0 ! M Age < 9.6 Gyr (H=50 km/s/Mpc)

179

Strategic Isotope Production | ornl.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

Strategic Isotope Strategic Isotope Production SHARE Strategic Isotope Production Typical capsules used in the transport of 252Cf source material inside heavily shielded shipping casks. ORNL's unique facilities at the High Flux Isotope Reactor (HFIR), Radiochemical Engineering Development Center (REDC), Irradiated Fuels Examination Laboratory (IFEL), and Irradiated Materials Examination Testing facility (IMET) are routinely used in the production, purification, packaging, and shipping of a number of isotopes of national importance, including: 75Se, 63Ni, 238Pu, 252Cf, and others. The intense neutron flux of the HFIR (2.0 x 1015 neutrons/cm²·s) permits the rapid formation of such isotopes. These highly irradiated materials are then processed and packaged for shipping using the facilities at the REDC, IFEL, and IMET.

180

Observation of the Isotopic Evolution of PWR Fuel Using an Antineutrino Detector  

E-Print Network (OSTI)

By operating an antineutrino detector of simple design during several fuel cycles, we have observed long term changes in antineutrino flux that result from the isotopic evolution of a commercial pressurized water reactor. Measurements made with simple antineutrino detectors of this kind offer an alternative means for verifying fissile inventories at reactors, as part of IAEA and other reactor safeguards regimes.

Bowden, N S; Dazeley, S; Svoboda, R; Misner, A; Palmer, T

2008-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

Precision spectroscopy with reactor anti-neutrinos  

E-Print Network (OSTI)

In this work we present an accurate parameterization of the anti-neutrino flux produced by the isotopes 235U, 239Pu and 241Pu in nuclear reactors. We determine the coefficients of this parameterization, as well as their covariance matrix, by performing a fit to spectra inferred from experimentally measured beta spectra. Subsequently we show that flux shape uncertainties play only a minor role in the KamLAND experiment, however, we find that future reactor neutrino experiments to measure the mixing angle $\\theta_{13}$ are sensitive to the fine details of the reactor neutrino spectra. Finally, we investigate the possibility to determine the isotopic composition in nuclear reactors through an anti-neutrino measurement. We find that with a 3 month exposure of a one ton detector the isotope fractions and the thermal reactor power can be determined at a few percent accuracy, which may open the possibility of an application for safeguard or non-proliferation objectives.

Huber, P; Huber, Patrick; Schwetz, Thomas

2004-01-01T23:59:59.000Z

182

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

TRAINING AND QUALIFICATION (TQ) TRAINING AND QUALIFICATION (TQ) OBJECTIVE TQ-1: The RRD selection, training, and qualification program for operations and operations support personnel has been established, documented, and implemented. The selection process and applicable position-specific training for managers ensures competence commensurate with responsibilities. (The training and qualification program encompasses the range of duties and activities that are required to be performed.) (Core Requirement 3) Criteria * The RRD training program has been defined, established, documented, and implemented. An assigned RRD staff member tracks the training requirements, including site-level training requirements, and completion by RRD personnel. * The selection process and implementation of position-specific training for

183

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Occupational Safety & Health Occupational Safety & Health OBJECTIVE ESH-1: Personnel exhibit an awareness of public and worker safety and health requirements and, through their actions, demonstrate a high-priority commitment to comply with these requirements. (Should also be coordinated with OP-1, MT-1, AB-2, EP-1, and ES-2) (CR-1) Criteria Personnel understand their right and responsibility to identify safety issues and invoke work suspension and stop work authority when necessary. Personnel anticipate, recognize, evaluate, and appropriately respond to hazards that may be present in the work place. Approach Record Review: Review the site policies and procedures which promote the identification and promulgation of safety concerns and work suspension and stop work

184

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

DOE Oversight DOE Oversight OBJECTIVE DOE-1: The DOE field office oversight programs, such as occurrence reporting, facility representative, corrective action, and quality assurance programs, are adequate. (Core Requirement 18) Criteria * The DOE field office oversight programs, such as occurrence reporting, facility representative, corrective action, and quality assurance programs, have been documented in ORO directives and are adequate. * The DOE oversight program is adequate and effective. Approach Record Review: Evaluate the assignments, responsibilities, and reporting relationships of the DOE Facility Representative(s) and program elements to ensure that Headquarters program direction is properly implemented. Review a sample of the Facility Representative's logs and ORION walkthrough reports.

185

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

MAINTENANCE AND TESTING (MT) MAINTENANCE AND TESTING (MT) OBJECTIVE MT-1: A program is in place to confirm and periodically reconfirm the condition and operability of safety SSCs. This includes examinations of test and calibration records of these systems. The material condition of all safety, process, and utility systems will support the safe conduct of work. (Core Requirement 8) Criteria * A test program is in place to confirm the condition and operability of safety SSCs for HFIR. * Testing is performed after maintenance activities to confirm the effectiveness of the maintenance and the operability of safety SSCs. * Surveillance testing is performed to maintain the operability of safety SSCs consistent with the TSR surveillance requirements. * The material condition of all safety, process, and utility systems will support

186

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

CONDUCT OF OPERATIONS (OP) CONDUCT OF OPERATIONS (OP) OBJECTIVE OP-1: The formality and discipline of operations are adequate to conduct work safely, and programs are in place to maintain this formality and discipline (e.g., DOE Order 5480.19). (Core Requirement 13) Criteria * The formality and discipline of operations are adequate to conduct work safely, and programs are in place to maintain this formality and discipline. * Implementation of ADM-0001, HFIR Conduct of Operations, is adequate for operations. * Implementation of ADM-0601, Cold Source Conduct of Operations, is adequate for operations. Approach Record Review: Review of ADM-0601 and ADM-0001 to confirm that those procedures fully capture the expectations for formality and discipline of operations.

187

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

AUTHORIZATION BASIS (AB) AUTHORIZATION BASIS (AB) OBJECTIVE AB-1: Facility safety documentation is in place and has been implemented that describes the safety envelope of the facility. The safety documentation should characterize the hazards/risks associated with the facility and should identify preventive and mitigating measures (such as systems, procedures, and administrative controls) that protect workers and the public from those hazards/risks. The SSCs are defined, and a system to maintain control over their design and modification is established. Adequate, correct procedures and safety limits are in place for operating the process and utility systems that include revisions for the modifications that have been made to the facility. (Core Requirements 7, 10, and 14)

188

A Proposal: Reliability Centered Maintenance (RCM) for the High Flux Isotope Reactor (HFIR)  

E-Print Network (OSTI)

#12;User Program and Community ­ User Statistics HFIR FY 2009HFIR FY 2009 Goal 300Goal 300SNS FY 2009SNS FY 2009 Goal 260Goal 260 358358 307307 250 300 350 400 queUsers 09 50 100 150 200 HFIR SNS 200 250 HFIR SNS niqueUsers 009 6060 3232 44 3030 11 6767 3737 11 2727 11 50 100 NumberofUn FY20 2

189

Meeting notes of the High Flux Isotope Reactor (HFIR) futures group  

SciTech Connect

This report is a compilation of the notes from the ten meetings. The group charter is: (1) to identify and characterize the range of possibilities and necessities for keeping the HFIR operating for at least the next 15 years; (2) to identify and characterize the range of possibilities for enhancing the scientific and technical utility of the HFIR; (3) to evaluate the benefits or impacts of these possibilities on the various scientific fields that use the HFIR or its products; (4) to evaluate the benefits or impacts on the operation and maintenance of the HFIR facility and the regulatory requirements; (5) to estimate the costs, including operating costs, and the schedules, including downtime, for these various possibilities; and one possible impact of proposed changes may be to stimulate increased pressure for a reduced enrichment fuel for HFIR.

Houser, M.M. [comp.

1995-08-01T23:59:59.000Z

190

ORNL - Restart of the High Flux Isotope Reactor 2-07  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

ENVIRONMENTAL PROTECTION AND WASTE MANAGEMENT (EW) ENVIRONMENTAL PROTECTION AND WASTE MANAGEMENT (EW) OBJECTIVE EW-1: UT-Battelle line management has established environmental protection and waste management programs to ensure safe accomplishment of work (or is adequately applying an existing, approved program). Personnel exhibit an awareness of environmental protection and waste management requirements, and through their actions, they demonstrate a high-priority commitment to comply with these requirements. (Core Requirements 1 and 14) Criteria * All environment compliance and waste management matrix support functions are identified for HFIR's operations. * Appropriate environmental protection/waste management plans and procedures for HFIR have been issued. * Adequate staffing is available to support the environmental protection and

191

Neutron flux measurements in the side-core region of Hunterston B advanced gas-cooled reactor  

Science Conference Proceedings (OSTI)

The core restraints of advanced gas-cooled reactors are important structural components that are required to maintain the geometric integrity of the cores. A review of neutron dosimetry for the sister stations Hunterston B and Hinkley Point B identified that earlier conservative assessments predicted high thermal neutron dose rates to key components of the restraint structure (the restraint rod welds), with the implication that some of them may be predicted to fail during a seismic event. A revised assessment was therefore undertaken [Thornton, D. A., Allen, D. A., Tyrrell, R. J., Meese, T. C., Huggon, A.P., Whiley, G. S., and Mossop, J. R., 'A Dosimetry Assessment for the Core Restraint of an Advanced Gas Cooled Reactor,' Proceedings of the 13. International Symposium on Reactor Dosimetry (ISRD-13, May 2008), World Scientific, River Edge, NJ, 2009, W. Voorbraak, L. Debarberis, and P. D'hondt, Eds., pp. 679-687] using a detailed 3D model and a Monte Carlo radiation transport program, MCBEND. This reassessment resulted in more realistic fast and thermal neutron dose recommendations, the latter in particular being much lower than had been thought previously. It is now desirable to improve confidence in these predictions by providing direct validation of the MCBEND model through the use of neutron flux measurements. This paper describes the programme of work being undertaken to deploy two neutron flux measurement 'stringers' within the side-core region of one of the Hunterston B reactors for the purpose of validating the MCBEND model. The design of the stringers and the determination of the preferred deployment locations have been informed by the use of detailed MCBEND flux calculations. These computational studies represent a rare opportunity to design a flux measurement beforehand, with the clear intention of minimising the anticipated uncertainties and obtaining measurements that are known to be representative of the neutron fields to which the vulnerable steel restraint components are exposed. (authors)

Allen, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Shaw, S.E. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom); Huggon, A.P.; Steadman, R.J.; Thornton, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Whiley, G.S. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom)

2011-07-01T23:59:59.000Z

192

Unsteady momentum fluxes in two-phase flow and the vibration of nuclear reactor components  

E-Print Network (OSTI)

The steady and unsteady components of the momentum flux in a twophase flow have been measured at the exit of a vertical pipe. Measured momentum flux data has been machine processed by standard random vibration techniques ...

Yih, Tien Sieh

1968-01-01T23:59:59.000Z

193

Inverse Beta Decay in a Nonequilibrium Antineutrino Flux from a Nuclear Reactor  

E-Print Network (OSTI)

The evolution of the reactor antineutrino spectrum toward equilibrium above the inverse beta-decay threshold during the reactor operating period and the decay of residual antineutrino radiation after reactor shutdown are considered. It is found that, under certain conditions, these processes can play a significant role in experiments seeking neutrino oscillations.

V. I. Kopeikin; L. A. Mikaelyan; V. V. Sinev

2001-10-23T23:59:59.000Z

194

1 Inverse Beta Decay in a Nonequilibrium Antineutrino Flux from a Nuclear Reactor  

E-Print Network (OSTI)

The evolution of the reactor antineutrino spectrum toward equilibrium above the inverse beta-decay threshold during the reactor operating period and the decay of residual ¯?e, radiation after reactor shutdown are considered. It is found that, under certain conditions, these processes can play a significant role in experiments seeking neutrino oscillations. 1.

V. I. Kopeikin A; L. A. Mikaelyan A; V. V. Sinev A

2001-01-01T23:59:59.000Z

195

Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor  

Science Conference Proceedings (OSTI)

This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

Ilas, Germina [ORNL; Gauld, Ian C [ORNL

2011-01-01T23:59:59.000Z

196

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

SciTech Connect

The High Flux Reactor (RHF) of the Laue Langevin Institute (ILL) based in Grenoble, France is a research reactor designed primarily for neutron beam experiments for fundamental science. It delivers one of the most intense neutron fluxes worldwide, with an unperturbed thermal neutron flux of 1.5 x 10{sup 15} n/cm{sup 2}/s in its reflector. The reactor has been conceived to operate at a nuclear power of 57 MW but currently operates at 52 MW. The reactor currently uses a Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most worldwide research and test reactors have already started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on a mixture of uranium and molybdenum (UMo) is expected to allow the conversion of compact high performance reactors like the RHF. This report presents the results of reactor design, performance and steady state safety analyses for conversion of the RHF from the use of HEU fuel to the use of UMo LEU fuel. The objective of this work was to show that is feasible, under a set of manufacturing assumptions, to design a new RHF fuel element that could safely replace the HEU element currently used. The new proposed design has been developed to maximize performance, minimize changes and preserve strong safety margins. Neutronics and thermal-hydraulics models of the RHF have been developed and qualified by benchmark against experiments and/or against other codes and models. The models developed were then used to evaluate the RHF performance if LEU UMo were to replace the current HEU fuel 'meat' without any geometric change to the fuel plates. Results of these direct replacement analyses have shown a significant degradation of the RHF performance, in terms of both neutron flux and cycle length. Consequently, ANL and ILL have collaborated to investigate alternative designs. A promising candidate design has been selected and studied, increasing the total amount of fuel without changing the external plate dimensions by relocating the burnable poison. In this way, changes required in the fuel element are reasonably small. With this new design, neutronics analyses have shown that performance could be maintained at a high level: 2 day decrease of cycle length (to 47.5 days at 58.3 MW) and 1-2% decrease of brightness in the cold and hot sources in comparison to the current typical operation. In addition, studies have shown that the thermal-hydraulic and shutdown margins for the proposed LEU design would satisfy technical specifications.

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

2010-08-19T23:59:59.000Z

197

Continuous production of tritium in an isotope-production reactor with a separate circulation system  

DOE Patents (OSTI)

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium is allowed to flow through the reactor in separate loops in order to facilitate the production and removal of tritium.

Cawley, W.E.; Omberg, R.P.

1982-08-19T23:59:59.000Z

198

Progress in the Use of Isotopes: The Atomic Triad - Reactors, Radioisotopes and Radiation  

DOE R&D Accomplishments (OSTI)

Recent years have seen a substantial growth in the use of isotopes in medicine, agriculture, and industry: up to the minute information on the production and use of isotopes in the U.S. is presented. The application of radioisotopes to industrial processes and manufacturing operations has expanded more rapidly than any one except its most ardent advocates expected. New uses and new users are numerous. The adoption by industry of low level counting techniques which make possible the use of carbon-14 and tritium in the control of industrial processes and in certain exploratory and research problems is perhaps most promising of current developments. The latest information on savings to industry will be presented. The medical application of isotopes has continued to develop at a rapid pace. The current trend appears to be in the direction of improvements in technique and the substitution of more effective isotopes for those presently in use. Potential and actual benefits accruing from the use of isotopes in agriculture are reviewed. The various methods of production of radioisotopes are discussed. Not only the present methods but also interesting new possibilities are covered. Although isotopes are but one of the many peaceful uses of the atom, it is the first to pay its way. (auth)

Libby, W. F.

1958-08-04T23:59:59.000Z

199

Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression  

DOE Patents (OSTI)

A high-power-density-laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems. 25 figs.

Lasche, G.P.

1987-02-20T23:59:59.000Z

200

Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression  

SciTech Connect

A high-power-density laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems.

Lasche, George P. (Arlington, VA)

1988-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Flux stability and power control in the Soviet RBMK-1000 reactors  

SciTech Connect

As a result of the Chernobyl accident, the Soviets have studied and implemented various design changes to improve the safety of the RBMK reactors. The safety measurements include modifications of the control rod configuration, fuel enrichment increase from 2.0 to 2.4 weight percent U-235, and installation of additional supplemental absorbers. The purpose of this study is to investigate the effects of increased fuel enrichment, different control rod positions, and supplemental absorber loadings on reactivity control, power distribution within the large RBMK core, and relative stability against power oscillations.

Meriwether, G.H.; McNeece, J.P.

1993-08-01T23:59:59.000Z

202

Vented target elements for use in an isotope-production reactor. [LMFBR  

DOE Patents (OSTI)

A method is described for producing tritium gas in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins equipped with vents, and tritium gas is recovered from the coolant.

Cawley, W.E.; Omberg, R.P.

1982-08-19T23:59:59.000Z

203

Assemblies with both target and fuel pins in an isotope-production reactor  

DOE Patents (OSTI)

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

Cawley, W.E.; Omberg, R.P.

1982-08-19T23:59:59.000Z

204

Fuel pins with both target and fuel pellets in an isotope-production reactor  

DOE Patents (OSTI)

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target pellets are placed in close contact with fissile fuel pellets in order to increase the tritium production rate.

Cawley, W.E.; Omberg, R.P.

1982-08-19T23:59:59.000Z

205

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

Science Conference Proceedings (OSTI)

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

Not Available

1992-07-01T23:59:59.000Z

206

Reactor Simulation for Antineutrino Experiments using DRAGON and MURE  

E-Print Network (OSTI)

Rising interest in nuclear reactors as a source of antineutrinos for experiments motivates validated, fast, and accessible simulations to predict reactor fission rates. Here we present results from the DRAGON and MURE simulation codes and compare them to other industry standards for reactor core modeling. We use published data from the Takahama-3 reactor to evaluate the quality of these simulations against the independently measured fuel isotopic composition. The propagation of the uncertainty in the reactor operating parameters to the resulting antineutrino flux predictions is also discussed.

Jones, C L; Conrad, J M; Djurcic, Z; Fallot, M; Giot, L; Keefer, G; Onillon, A; Winslow, L

2011-01-01T23:59:59.000Z

207

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors  

Science Conference Proceedings (OSTI)

The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

Rosenthal, Murray Wilford [ORNL

2009-08-01T23:59:59.000Z

208

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

Science Conference Proceedings (OSTI)

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.

Evan Harpeneau

2011-06-24T23:59:59.000Z

209

Design of a low enrichment, enhanced fast flux core for the Massachusetts Institute of Technology Research Reactor  

E-Print Network (OSTI)

Worldwide, there is limited test reactor capacity to perform the required irradiation experiments on advanced fast reactor materials and fuel designs. This is particularly true in the U.S., which no longer has an operating ...

Ellis, Tyler Shawn

2009-01-01T23:59:59.000Z

210

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

SciTech Connect

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

E.M. Harpenau

2010-11-15T23:59:59.000Z

211

A Brief History i-l Research Reactors  

E-Print Network (OSTI)

stainless steel sam- ples in the High Flux Isotope Reactor (HFIR) at tem- peratures of 380 to 680" with up/cm' to balance the gas pressure were used m their calculation. A comparison of the results with HFIR and the HFIR ex- perimental data is presented in section 5. Applications of the model to various fusion designs

212

Modernization of the High Flux Isotope Reactor (HFIR) to Provide a Cold Neutron Source and Experimentation Facility  

Science Conference Proceedings (OSTI)

This paper discusses the installation of a cold neutron source at HFIR with respect to the project as a modernization of the facility. The paper focuses on why the project was required, the scope of the cold source project with specific emphasis on the design, and project management information.

Rothrock, Benjamin G [ORNL; Farrar, Mike B [ORNL

2009-01-01T23:59:59.000Z

213

Intercomparison of the finite difference and nodal discrete ordinates and surface flux transport methods for a LWR pool-reactor benchmark problem in X-Y geometry  

Science Conference Proceedings (OSTI)

The aim of the present work is to compare and discuss the three of the most advanced two dimensional transport methods, the finite difference and nodal discrete ordinates and surface flux method, incorporated into the transport codes TWODANT, TWOTRAN-NODAL, MULTIMEDIUM and SURCU. For intercomparison the eigenvalue and the neutron flux distribution are calculated using these codes in the LWR pool reactor benchmark problem. Additionally the results are compared with some results obtained by French collision probability transport codes MARSYAS and TRIDENT. Because the transport solution of this benchmark problem is close to its diffusion solution some results obtained by the finite element diffusion code FINELM and the finite difference diffusion code DIFF-2D are included.

O'Dell, R.D.; Stepanek, J.; Wagner, M.R.

1983-01-01T23:59:59.000Z

214

REACTOR  

DOE Patents (OSTI)

A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

Roman, W.G.

1961-06-27T23:59:59.000Z

215

ISOTOPE CONVERSION DEVICE  

DOE Patents (OSTI)

This patent relates to nuclear reactors of tbe type utilizing a liquid fuel and designed to convert a non-thermally fissionable isotope to a thermally fissionable isotope by neutron absorption. A tank containing a reactive composition of a thermally fissionable isotope dispersed in a liquid moderator is disposed within an outer tank containing a slurry of a non-thermally fissionable isotope convertible to a thermally fissionable isotope by neutron absorption. A control rod is used to control the chain reaction in the reactive composition and means are provided for circulating and cooling the reactive composition and slurry in separate circuits.

Wigner, E.P.; Young, G.J.; Ohlinger, L.A.

1957-12-01T23:59:59.000Z

216

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

SciTech Connect

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.

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

1990-05-01T23:59:59.000Z

217

CALCULATION OF EFFECT OF FUEL BURNUP ON FUEL AND POISON DISTRIBUTIONS AND ON FLUX DISTRIBUTION IN THE MARINE REACTOR  

SciTech Connect

The changes in fuel and poison distributions and the changes in flux shape which accompany the consumption of fuel are studied. The technique employed is a perturbation calculation based on a one-velocity group treatment of the neutrons. The geometry is a spherical core surrounded by an infinite reflector. The programming forms for the IBM-650 which performs the computations using the Wolontis interpretative system are included. Two sample calculations were worked out using the code described and the results are plotted. (M.H.R.)

Hinman, G.

1957-01-29T23:59:59.000Z

218

The BURNUP package of applied programs used for computing the isotopic composition of materials of an operating nuclear reactor  

SciTech Connect

This paper described the procedure of implementation and the possibilities of the BURNUP program. The purpose of the program is to predict the change in the nuclear composition of the materials of which a reactor is made in the course of its run and compute the radiation characteristics of the materials after their irradiation.

Yudkevich, M. S., E-mail: umark@adis.vver.kiae.ru [National Research Centre Kurchatov Institute (Russian Federation)

2012-12-15T23:59:59.000Z

219

Fast Flux Test Facility (FFTF) Briefing Book 1 Summary  

SciTech Connect

This report documents the results of evaluations preformed during 1997 to determine what, if an, future role the Fast Flux Test Facility (FFTF) might have in support of the Department of Energy’s tritium productions strategy. An evaluation was also conducted to assess the potential for the FFTF to produce medical isotopes. No safety, environmental, or technical issues associated with producing 1.5 kilograms of tritium per year in the FFTF have been identified that would change the previous evaluations by the Department of Energy, the JASON panel, or Putnam, Hayes & Bartlett. The FFTF can be refitted and restated by July 2002 for a total expenditure of $371 million, with an additional $64 million of startup expense necessary to incorporate the production of medical isotopes. Therapeutic and diagnostic applications of reactor-generated medical isotopes will increase dramatically over the next decade. Essential medical isotopes can be produced in the FFTF simultaneously with tritium production, and while a stand-alone medical isotope mission for the facility cannot be economically justified given current marker conditions, conservative estimates based on a report by Frost &Sullivan indicate that 60% of the annual operational costs (reactor and fuel supply) could be offset by revenues from medical isotope production within 10 yeas of restart. The recommendation of the report is for the Department of Energy to continue to maintain the FFTF in standby and proceed with preparation of appropriate Nations Environmental Policy Act documentation in full consultation with the public to consider the FFTF as an interim tritium production option (1.5 kilograms/year) with a secondary mission of producing medical isotopes.

WJ Apley

1997-12-01T23:59:59.000Z

220

CONTROL MEANS FOR REACTOR  

DOE Patents (OSTI)

An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

Manley, J.H.

1961-06-27T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

Controls on Fluxes of Mercury in Aquatic Food Webs: Application of the Dynamic Mercury Cycling Model to Four Enclosure Experiments w ith Additions of Stable Mercury Isotopes  

Science Conference Proceedings (OSTI)

New controls on utility mercury emissions are under consideration in order to limit human exposure to mercury resulting from fish consumption. Evaluation of such measures requires an understanding of how mercury cycles through lakes and streams. This report describes the application of EPRI's Dynamic Mercury Cycling Model (D-MCM) to experiments involving the addition of stable mercury Hg(II) isotopes to four 10-meter-diameter enclosures in a lake.

2001-09-21T23:59:59.000Z

222

The influence of helium on mechanical properties of model austenitic alloys, determined using sup 59 Ni isotopic tailoring and fast reactor irradiation  

Science Conference Proceedings (OSTI)

The objective of this effort is to study the separate and synergistic effects of helium and other important variables on the evolution of microstructure and macroscopic properties during irradiation of structural metals. The alloys employed in this study were nominally Fe-15Cr-25Ni, Fe-15Cr-25Ni-0.04P and Fe-15Cr-45Ni (wt %) in both the cold worked and annealed conditions. Tensile testing and microscopy continue on specimens removed from the first, second and third discharges of the {sup 59}Ni isotopic doping experiment. The results to date indicate that helium/dpa ratios typical of fusion reactors (4 to 19 appm/dpa) do not lead to significant changes in the yield strength of model Fe-Cr-Ni alloys. Measurements of helium generated in undoped specimens from the second and third discharges show that the helium/dpa ratio increases during irradiation in FFTF due to the production of {sup 59}Ni. In specimens doped with {sup 59}Ni prior to irradiation, the helium/dpa ratio can increase, decrease or remain the same during the second irradiation interval. This behavior occurs because the cross sections for the production and burnout of {sup 59}Ni are very sensitive to core location and the nature of neighboring components. 14 refs., 5 figs., 3 tabs.

Hamilton, M.L.; Garner, F.A. (Pacific Northwest Lab., Richland, WA (USA)); Oliver, B.M. (Rockwell International Corp., Canoga Park, CA (USA))

1990-11-01T23:59:59.000Z

223

NEUTRON FLUX INTENSITY DETECTION  

DOE Patents (OSTI)

A method of measuring the instantaneous intensity of neutron flux in the core of a nuclear reactor is described. A target gas capable of being transmuted by neutron bombardment to a product having a resonance absorption line nt a particular microwave frequency is passed through the core of the reactor. Frequency-modulated microwave energy is passed through the target gas and the attenuation of the energy due to the formation of the transmuted product is measured. (AEC)

Russell, J.T.

1964-04-21T23:59:59.000Z

224

POWER REACTOR  

DOE Patents (OSTI)

A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

Zinn, W.H.

1958-07-01T23:59:59.000Z

225

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

Christy, R.F.

1958-07-15T23:59:59.000Z

226

Production capabilities in US nuclear reactors for medical radioisotopes  

SciTech Connect

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.

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)

1992-11-01T23:59:59.000Z

227

Neutronic Study of Slightly Modified Water Reactors and Application to Transition Scenarios  

SciTech Connect

In this paper we have studied slightly modified water reactors and their applications to transition scenarios. The PWR and CANDU reactors have been considered. New fuels based on Thorium have been tested: Thorium/Plutonium and Thorium/Uranium- 233, with different fissile isotope contents. Changes in the geometry of the assemblies were also explored to modify the moderation ratio, and consequently the neutron flux spectrum. A core equivalent assembly methodology was introduced as an exploratory approach and to reduce the computation time. Several basic safety analyses were also performed. We have finally developed a new scenario code, named OSCAR (Optimized Scenario Code for Advanced Reactors), to study the efficiency of these modified reactors in transition to Gen IV reactors or in symbiotic fleet. (authors)

Chambon, Richard; Guillemin, Perrine; Nuttin, Alexis; Bidaud, A. [LPSC, Universite Joseph Fourier Grenoble 1, CNRS/IN2P3, Institut National Polytechnique de Grenoble 53 Av. Des Martyrs, 38000 Grenoble (France); Capellan, N.; David, S.; Meplan, O.; Wilson, J. [Institut de Physique Nucleaire - IPN, 15 rue Georges Clemenceau 91406 Orsay (France)

2007-07-01T23:59:59.000Z

228

Safety and core design of large liquid-metal cooled fast breeder reactors  

E-Print Network (OSTI)

well known from basic reactor theory, the flux distributionof a fast reactor using the perturbation theory”. In: Atomicbeam theory and are not specific to a nuclear reactor core.

Qvist, Staffan Alexander

2013-01-01T23:59:59.000Z

229

COMSOL-based Nuclear Reactor Kinetics Studies at the HFIR  

Science Conference Proceedings (OSTI)

The computational ability to accurately predict the dynamic behavior of a nuclear reactor core in response to reactivity-induced perturbations is an important subject in reactor physics. Space-time and point kinetics methodologies were developed for the purpose of studying the transient-induced behavior of the High Flux Isotope Reactor s (HFIR) compact core. The space-time simulations employed the three-energy-group neutron diffusion equations, and transients initiated by control cylinder and hydraulic tube rabbit ejections were studied. The work presented here is the first step towards creating a comprehensive multiphysics methodology for studying the dynamic behavior of the HFIR core during reactivity perturbations. The results of these studies show that point kinetics is adequate for small perturbations in which the power distribution is assumed to be time-independent, but space-time methods must be utilized to determine localized effects.

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

2011-01-01T23:59:59.000Z

230

Reactor operations informal report, October 1994  

Science Conference Proceedings (OSTI)

This monthly progress report is divided into two parts. Part one covers the Brookhaven Medical Research Reactor and part two covers the Brookhaven High Flux Beam Reactor. Information is given for each reactor covering the following areas: reactor operation; instrumentation; mechanical maintenance; occurrence reports; and reactor safety.

Hauptman, H.M.; Petro, J.N.; Jacobi, O.; Lettieri, V.; Holden, N.; Ports, D.; Petricek, R.

1994-10-01T23:59:59.000Z

231

Development of Technical Nuclear Forensics for Spent Research Reactor Fuel  

E-Print Network (OSTI)

Pre-detonation technical nuclear forensics techniques for research reactor spent fuel were developed in a collaborative project with Savannah River National Lab ratory. An inverse analysis method was employed to reconstruct reactor parameters from a spent fuel sample using results from a radiochemical analysis. In the inverse analysis, a reactor physics code is used as a forward model. Verification and validation of different reactor physics codes was performed for usage in the inverse analysis. The verification and validation process consisted of two parts. The first is a variance analysis of Monte Carlo reactor physics burnup simulation results. The codes used in this work are MONTEBURNS and MCNPX/CINDER. Both utilize Monte Carlo transport calculations for reaction rate and flux results. Neither code has a variance analysis that will propagate through depletion steps, so a method to quantify and understand the variance propagation through these depletion calculations was developed. The second verification and validation process consisted of comparing reactor physics code output isotopic compositions to radiochemical analysis results. A sample from an Oak Ridge Research Reactor spent fuel assembly was acquired through a drilling process. This sample was then dissolved in nitric acid and diluted in three different quantities, creating three separate samples. A radiochemical analysis was completed and the results were compared to simulation outputs at different levels ofdetail. After establishing a forward model, an inverse analysis was developed to re-construct the burnup, initial uranium isotopic compositions, and cooling time of a research reactor spent fuel sample. A convergence acceleration technique was used that consisted of an analytical calculation to predict burnup, initial 235U, and 236U enrichments. The analytic calculation results may also be used stand alone or in a database search algorithm. In this work, a reactor physics code is used as a for- ward model with the analytic results as initial conditions in a numerical optimization algorithm. In the numerical analysis, the burnup and initial uranium isotopic com- positions are reconstructed until the iterative spent fuel characteristics converge with the measured data. Upon convergence of the sample’s burnup and initial uranium isotopic composition, the cooling time can be reconstructed. To reconstruct cooling time, the standard decay equation is inverted and solved for time. Two methods were developed. One method uses the converged burnup and initial uranium isotopic compositions along in a reactor depletion simulation. The second method uses an isotopic signature that does not decay out of its mass bin and has a simple production chain. An example would be 137Cs which decays into the stable 137Ba. Similar results are achieved with both methods, but extended shutdown time or time away from power results in over prediction of the cooling time. The over prediction of cooling time and comparison of different burnup reconstruction isotope results are indicator signatures of extended shutdown or time away from power. Due to dynamic operation in time and function, detailed power history reconstruction for research reactors is very challenging. Frequent variations in power, repeated variable shutdown time length, and experimentation history affect the spectrum an individual assembly is burned with such that full reactor parameter reconstruction is difficult. The results from this technical nuclear forensic analysis may be used with law enforcement, intelligence data, macroscopic and microscopic sample characteristics in a process called attribution to suggest or exclude possible sources of origin for a sample.

Sternat, Matthew 1982-

2012-12-01T23:59:59.000Z

232

Oxygen Isotopes  

NLE Websites -- All DOE Office Websites (Extended Search)

Pages to Isotopes Data Modern Records of Carbon and Oxygen Isotopes in Atmospheric Carbon Dioxide and Carbon-13 in Methane 800,000 Deuterium Record and Shorter Records of...

233

NUCLEAR REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

Currier, E.L. Jr.; Nicklas, J.H.

1963-06-11T23:59:59.000Z

234

Risk management at the Oak Ridge National Laboratory research reactors  

SciTech Connect

In November of 1986, the High Flux Isotope Reactor (HFIR) was shut down by Oak Ridge National Laboratory (ORNL) due to a concern regarding embrittlement of the reactor vessel. A massive review effort was undertaken by ORNL and the Department of Energy (DOE). This review resulted in an extensive list of analyses and design modifications to be completed before restart could take place. The review also focused on the improvement of management practices including implementation of several of the Institute of Nuclear Power Operations (INPO) requirements. One of the early items identified was the need to perform a Probabilistic Risk Assessment (PRA) on the reactor. It was decided by ORNL management that this PRA would not be just an exercise to assess the ``bottom`` line in order to restart, but would be used to improve the overall safety of the reactor, especially since resources (both manpower and dollars) were severely limited. The PRA would become a basic safety tool to be used instead of a more standard deterministic approach to safety used in commercial reactor power plants. This approach was further reinforced, because the reactor was nearly 25 years old at this time, and the design standards and regulations had changed significantly since the original design, and many of the safety issues could not be addressed by compliance to codes and standards.

Flanagan, G.F.; Linn, M.A.; Proctor, L.D.; Cook, D.H.

1994-12-31T23:59:59.000Z

235

THERMAL NEUTRONIC REACTOR  

DOE Patents (OSTI)

A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

Spinrad, B.I.

1960-01-12T23:59:59.000Z

236

Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors  

DOE Patents (OSTI)

A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify /sup 233/U, /sup 235/U and /sup 239/Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as /sup 240/Pu, /sup 244/Cm and /sup 252/Cf, and the spontaneous alpha particle emitter /sup 241/Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether permanent low-level burial is appropriate for the waste sample.

Caldwell, J.T.; Kunz, W.E.; Atencio, J.D.

1982-03-31T23:59:59.000Z

237

Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors  

DOE Patents (OSTI)

A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify .sup.233 U, .sup.235 U and .sup.239 Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as .sup.240 Pu, .sup.244 Cm and .sup.252 Cf, and the spontaneous alpha particle emitter .sup.241 Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether "permanent" low-level burial is appropriate for the waste sample.

Caldwell, John T. (Los Alamos, NM); Kunz, Walter E. (Santa Fe, NM); Atencio, James D. (Los Alamos, NM)

1984-01-01T23:59:59.000Z

238

Carbon Isotopes  

NLE Websites -- All DOE Office Websites (Extended Search)

Atmospheric Trace Gases » Carbon Isotopes Atmospheric Trace Gases » Carbon Isotopes Carbon Isotopes Gateway Pages to Isotopes Data Modern Records of Carbon and Oxygen Isotopes in Atmospheric Carbon Dioxide and Carbon-13 in Methane 800,000 Deuterium Record and Shorter Records of Various Isotopic Species from Ice Cores Carbon-13 13C in CO Measurements from Niwot Ridge, Colorado and Montana de Oro, California (Tyler) 13C in CO2 NOAA/CMDL Flask Network (White and Vaughn) CSIRO GASLAB Flask Network (Allison, Francey, and Krummel) CSIRO in situ measurements at Cape Grim, Tasmania (Francey and Allison) Scripps Institution of Oceanography (Keeling et al.) 13C in CH4 NOAA/CMDL Flask Network (Miller and White) Northern & Southern Hemisphere Sites (Quay and Stutsman) Northern & Southern Hemisphere Sites (Stevens)

239

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

SciTech Connect

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.

McBee, M.R. (ed.); Axe, J.D.; Hayter, J.B.

1990-07-01T23:59:59.000Z

240

Fission reactors and materials  

SciTech Connect

The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions.

Frost, B.R.T.

1981-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

RBU: A COMBINED MONTE CARLO REACTOR-BURNUP PROGRAM FOR THE IBM 709  

SciTech Connect

RBU is a digital computer program for the detailed calculation of the neutron, reactivity, and isotopic history of a reactor in which relatively exact models of the geometry and physical processes are included to permit reliable pre dictions of fuel costs and reactor performance. The program uses the Monte Carlo method to obtain the fine structure of the neutron flux in three space dimensions and energy. Using this fine structure, cross sections are averaged over space and energy to obtain the neutronic properties for equivalent homogeneous one- dimensional regions of space and ranges of energy. These are used in diffusion calculations to obtain the macroscopic flux distribution throughout the reactor. The consumption and production of isotopes is computed for a time step by the solution of sets of partial differential equations involving both the macroscopic and microscopic fluxes. With the new concentrations, diffusion calculations are performed again to obtain macroscopic fluxes for the next time step. At variable intervals, Monte Carlo calculations are again performed to determine the changes in microscopic flux distributions. The cycle is repeated until conditions on the reactivity or other properties dictate the end of the calculation. Programmed control rod manipulation may be included in the calculation. The Monte Carlo, diffusion, or burnup portions of the program may be used separately. The unresolved resonance range is treated by random selection of resonance parameters from appropriate distributions using the Doppler broadened single level Breit- Wigner formula. Resolved resonances are treated similarly with the exception that specific values of the resonance parameters are used. The effects of molecular binding and thermal motion of the nuclei on near-thermal scattering are treated by a simple model capable of incorporating the pertinent physical theory and data. (auth)

Leshan, E.J.; Burr, J.R.; Temme, M.; Thompson, G.T.; Triplett, J.R.

1959-09-30T23:59:59.000Z

242

Accurate determination of Curium and Californium isotopic ratios by inductively coupled plasma quadrupole mass spectrometry (ICP-QMS) in 248Cm samples for transmutation studies  

Science Conference Proceedings (OSTI)

The French Atomic Energy Commission has carried out several experiments including the mini-INCA (INcineration of Actinides) project for the study of minor-actinide transmutation processes in high intensity thermal neutron fluxes, in view of proposing solutions to reduce the radiotoxicity of long-lived nuclear wastes. In this context, a Cm sample enriched in {sup 248}Cm ({approx}97 %) was irradiated in thermal neutron flux at the High Flux Reactor (HFR) of the Laue-Langevin Institute (ILL). This work describes a quadrupole ICP-MS (ICP-QMS) analytical procedure for precise and accurate isotopic composition determination of Cm before sample irradiation and of Cm and Cf after sample irradiation. The factors that affect the accuracy and reproducibility of isotopic ratio measurements by ICP-QMS, such as peak centre correction, detector dead time, mass bias, abundance sensitivity and hydrides formation, instrumental background, and memory blank were carefully evaluated and corrected. Uncertainties of the isotopic ratios, taking into account internal precision of isotope ratio measurements, peak tailing, and hydrides formations ranged from 0.3% to 1.3%. This uncertainties range is quite acceptable for the nuclear data to be used in transmutation studies.

Gourgiotis, A.; Isnard, H.; Aubert, M.; Dupont, E.; AlMahamid, I.; Cassette, P.; Panebianco, S.; Letourneau, A.; Chartier, F.; Tian, G.; Rao, L.; Lukens, W.

2011-02-01T23:59:59.000Z

243

Recovery and Packaging of Tritium from Canadian Heavy Water Reactors  

Science Conference Proceedings (OSTI)

Fission Reactor / Proceedings of the Second National Topical Meeting on Tritium Technology in Fission, Fusion and Isotopic Applications (Dayton, Ohio, April 30 to May 2, 1985)

W.J. Holtslander; T.E. Harrison; V. Goyette; J.M. Miller

244

CRAD, Conduct of Operations - Oak Ridge National Laboratory High...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

High Flux Isotope Reactor Contractor ORR CRAD, Conduct of Operations - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February 2007 A section of Appendix C...

245

CRAD, Configuration Management - Oak Ridge National Laboratory...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor February...

246

CRAD, Emergency Management - Oak Ridge National Laboratory High...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor February 2007 A...

247

CRAD, Emergency Management - Oak Ridge National Laboratory High...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Emergency Management - Oak Ridge National Laboratory High Flux Isotope Reactor...

248

Chromatographic hydrogen isotope separation  

DOE Patents (OSTI)

Intermetallic compounds with the CaCu.sub.5 type of crystal structure, particularly LaNiCo.sub.4 and CaNi.sub.5, exhibit high separation factors and fast equilibrium times and therefore are useful for packing a chromatographic hydrogen isotope separation colum. The addition of an inert metal to dilute the hydride improves performance of the column. A large scale mutli-stage chromatographic separation process run as a secondary process off a hydrogen feedstream from an industrial plant which uses large volumes of hydrogen can produce large quantities of heavy water at an effective cost for use in heavy water reactors.

Aldridge, Frederick T. (Livermore, CA)

1981-01-01T23:59:59.000Z

249

Final Report, NEAC Subcommittee for Isotope Research & Production Planning  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Final Report, NEAC Subcommittee for Isotope Research & Production Final Report, NEAC Subcommittee for Isotope Research & Production Planning Final Report, NEAC Subcommittee for Isotope Research & Production Planning Isotopes, including both radioactive and stable isotopes, make important contributions to research, medicine, and industry in the United States and throughout the world. For nearly fifty years, the Department of Energy (DOE) has actively promoted the use of isotopes by funding (a) production of isotopes at a number of national laboratories with unique nuclear reactors or particle accelerators, (b) nuclear medicine research at the laboratories and in academia, (c) research into industrial applications of isotopes, and (d) research into isotope production and processing methods. The radio- pharmaceutical and radiopharmacy industries have their origin in

250

from Isotope Production Facility  

NLE Websites -- All DOE Office Websites (Extended Search)

Cancer-fighting treatment gets boost from Isotope Production Facility April 13, 2012 Isotope Production Facility produces cancer-fighting actinium - 2 - 2:32 Isotope cancer...

251

Laser Isotope Enrichment for Medical and Industrial Applications  

SciTech Connect

Laser Isotope Enrichment for Medical and Industrial Applications by Jeff Eerkens (University of Missouri), Jay Kunze (Idaho State University), and Leonard Bond (Idaho National Laboratory) The principal isotope enrichment business in the world is the enrichment of uranium for commercial power reactor fuels. However, there are a number of other needs for separated isotopes. Some examples are: 1) Pure isotopic targets for irradiation to produce medical radioisotopes. 2) Pure isotopes for semiconductors. 3) Low neutron capture isotopes for various uses in nuclear reactors. 4) Isotopes for industrial tracer/identification applications. Examples of interest to medicine are targets to produce radio-isotopes such as S-33, Mo-98, Mo-100, W-186, Sn-112; while for MRI diagnostics, the non-radioactive Xe-129 isotope is wanted. For super-semiconductor applications some desired industrial isotopes are Si-28, Ga-69, Ge-74, Se-80, Te-128, etc. An example of a low cross section isotope for use in reactors is Zn-68 as a corrosion inhibitor material in nuclear reactor primary systems. Neutron activation of Ar isotopes is of interest in industrial tracer and diagnostic applications (e.g. oil-logging). . In the past few years there has been a sufficient supply of isotopes in common demand, because of huge Russian stockpiles produced with old electromagnetic and centrifuge separators previously used for uranium enrichment. Production of specialized isotopes in the USA has been largely accomplished using old ”calutrons” (electromagnetic separators) at Oak Ridge National Laboratory. These methods of separating isotopes are rather energy inefficient. Use of lasers for isotope separation has been considered for many decades. None of the proposed methods have attained sufficient proof of principal status to be economically attractive to pursue commercially. Some of the authors have succeeded in separating sulfur isotopes using a rather new and different method, known as condensation repression. In this scheme a gas, of the selected isotopes for enrichment, is irradiated with a laser at a particular wavelength that would excite only one of the isotopes. The entire gas is subject to low temperatures sufficient to cause condensation on a cold surface. Those molecules in the gas that the laser excited are not as likely to condense as are the unexcited molecules. Hence the gas drawn out of the system will be enriched in the isotope that was excited by the laser. We have evaluated the relative energy required in this process if applied on a commercial scale. We estimate the energy required for laser isotope enrichment is about 20% of that required in centrifuge separations, and 2% of that required by use of "calutrons".

Leonard Bond

2006-07-01T23:59:59.000Z

252

Decommissioning of the High Flux Beam Reactor  

NLE Websites -- All DOE Office Websites (Extended Search)

After careful planning and input from regulatory agencies and the community, a decommissioning plan for the HFBR has been finalized. A Feasibility Study was completed and a...

253

High Flux Beam Reactor | Environmental Restoration Projects ...  

NLE Websites -- All DOE Office Websites (Extended Search)

been taken to partially decommission and prepare the HFBR for safe storage. Final decommissioning of the HFBR building will be performed at the completion of the decay period. The...

254

High Flux Beam Reactor | Environmental Restoration Projects ...  

NLE Websites -- All DOE Office Websites (Extended Search)

Reports HFBR Waste Loading Area, Soil Remediation (PDF) - July 2009 HFBR Decommissioning Project, Removal of the Control Rod Blades and Beam Plugs (PDF) - January 2010...

255

Apparatus for isotopic alteration of mercury vapor  

DOE Patents (OSTI)

An apparatus for enriching the isotopic Hg content of mercury is provided. The apparatus includes a reactor, a low pressure electric discharge lamp containing a fill including mercury and an inert gas. A filter is arranged concentrically around the lamp. In a preferred embodiment, constant mercury pressure is maintained in the filter by means of a water-cooled tube that depends from it, the tube having a drop of mercury disposed in it. The reactor is arranged around the filter, whereby radiation from said lamp passes through the filter and into said reactor. The lamp, the filter and the reactor are formed of a material which is transparent to ultraviolet light.

Grossman, Mark W. (Belmont, MA); George, William A. (Gloucester, MA); Marcucci, Rudolph V. (Danvers, MA)

1988-01-01T23:59:59.000Z

256

Producing tritium in a homogenous reactor  

DOE Patents (OSTI)

A method and apparatus are described for the joint production and separation of tritium. Tritium is produced in an aqueous homogenous reactor and heat from the nuclear reaction is used to distill tritium from the lower isotopes of hydrogen.

Cawley, William E. (Richland, WA)

1985-01-01T23:59:59.000Z

257

ISOTOPE SEPARATORS  

DOE Patents (OSTI)

An improvement is presented in the structure of an isotope separation apparatus and, in particular, is concerned with a magnetically operated shutter associated with a window which is provided for the purpose of enabling the operator to view the processes going on within the interior of the apparatus. The shutier is mounted to close under the force of gravity in the absence of any other force. By closing an electrical circuit to a coil mouated on the shutter the magnetic field of the isotope separating apparatus coacts with the magnetic field of the coil to force the shutter to the open position.

Bacon, C.G.

1958-08-26T23:59:59.000Z

258

Nuclear Archeology for CANDU Power Reactors  

SciTech Connect

The goal of this work is the development of so-called 'nuclear archeology' techniques to predict the irradiation history of both fuel-related and non-fuel-related materials irradiated in the CANDU (CANada Deuterium Uranium) family of nuclear reactors. In this application to CANDU-type reactors, two different scenarios for the collection of the appropriate data for use in these procedures will be assumed: the first scenario is the removal of the pressure tubes, calandria tubes, or fuel cladding and destructive analysis of the activation products contained in these structural materials; the second scenario is the nondestructive analysis (NDA) of the same hardware items via high-resolution gamma ray scans. There are obvious advantages and disadvantages for each approach; however, the NDA approach is the central focus of this work because of its simplicity and lack of invasiveness. The use of these techniques along with a previously developed inverse capability is expected to allow for the prediction of average flux levels and irradiation time, and the total fluence for samples where the values of selected isotopes can be measured.

Broadhead, Bryan L [ORNL

2011-01-01T23:59:59.000Z

259

Selected Isotopes for Optimized Fuel Assembly Tags  

SciTech Connect

In support of our ongoing signatures project we present information on 3 isotopes selected for possible application in optimized tags that could be applied to fuel assemblies to provide an objective measure of burnup. 1. Important factors for an optimized tag are compatibility with the reactor environment (corrosion resistance), low radioactive activation, at least 2 stable isotopes, moderate neutron absorption cross-section, which gives significant changes in isotope ratios over typical fuel assembly irradiation levels, and ease of measurement in the SIMS machine 2. From the candidate isotopes presented in the 3rd FY 08 Quarterly Report, the most promising appear to be Titanium, Hafnium, and Platinum. The other candidate isotopes (Iron, Tungsten, exhibited inadequate corrosion resistance and/or had neutron capture cross-sections either too high or too low for the burnup range of interest.

Gerlach, David C.; Mitchell, Mark R.; Reid, Bruce D.; Gesh, Christopher J.; Hurley, David E.

2008-10-01T23:59:59.000Z

260

Fluid-Structure Interaction for Coolant Flow in Research-type Nuclear Reactors  

Science Conference Proceedings (OSTI)

The High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), is scheduled to undergo a conversion of the fuel used and this proposed change requires an extensive analysis of the flow through the reactor core. The core consists of 540 very thin and long fuel plates through which the coolant (water) flows at a very high rate. Therefore, the design and the flow conditions make the plates prone to dynamic and static deflections, which may result in flow blockage and structural failure which in turn may cause core damage. To investigate the coolant flow between fuel plates and associated structural deflections, the Fluid-Structure Interaction (FSI) module in COMSOL will be used. Flow induced flutter and static deflections will be examined. To verify the FSI module, a test case of a cylinder in crossflow, with vortex induced vibrations was performed and validated.

Curtis, Franklin G [ORNL; Ekici, Kivanc [ORNL; Freels, James D [ORNL

2011-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Continuous Reactors Measure some of the neutrons all of the ...  

Science Conference Proceedings (OSTI)

... reactors use compact cores and highly enriched fuel (over ... U-235) in order to achieve high neutron fluxes ... the use of intermediate enrichment (20-50 ...

2009-11-29T23:59:59.000Z

262

Apparatus for an Inertial Fusion Reactor Inventor Abraham Massry...  

NLE Websites -- All DOE Office Websites (Extended Search)

Apparatus for an Inertial Fusion Reactor Inventor Abraham Massry This invention is comprised of a very large vacuum chamber capable of withstanding a very high neutron flux...

263

Neutron beam characterization at the Neutron Radiography Reactor (NRAD)  

Science Conference Proceedings (OSTI)

The Neutron Radiography Reactor (NRAD) is a 250-kW TRIGA Reactor operated by Argonne National Laboratory and is located near Idaho Falls, Idaho. The reactor and its facilities regarding radiography are detailed in another paper at this conference; this paper summarizes neutron flux measurements and calculations that have been performed to better understand and potentially improve the neutronics characteristics of the reactor.

Imel, G.R.; Urbatsch, T.; Pruett, D.P.; Ross, J.R.

1990-01-01T23:59:59.000Z

264

Advanced Materials at ORNL - Oak Ridge National Laboratory  

Nuclear materials Strategy: Take advantage of •Unique capabilities in synthesis, characterization, theory ... High Flux Isotope Reactor:

265

Fast flux locked loop  

DOE Patents (OSTI)

A flux locked loop for providing an electrical feedback signal, the flux locked loop employing radio-frequency components and technology to extend the flux modulation frequency and tracking loop bandwidth. The flux locked loop of the present invention has particularly useful application in read-out electronics for DC SQUID magnetic measurement systems, in which case the electrical signal output by the flux locked loop represents an unknown magnetic flux applied to the DC SQUID.

Ganther, Jr., Kenneth R. (Olathe, KS); Snapp, Lowell D. (Independence, MO)

2002-09-10T23:59:59.000Z

266

Radioactive isotope production for medical applications using Kharkov electron driven subcritical assembly facility.  

SciTech Connect

Kharkov Institute of Physics and Technology (KIPT) of Ukraine has a plan to construct an accelerator driven subcritical assembly. The main functions of the subcritical assembly are the medical isotope production, neutron thereby, and the support of the Ukraine nuclear industry. Reactor physics experiments and material research will be carried out using the capabilities of this facility. The United States of America and Ukraine have started collaboration activity for developing a conceptual design for this facility with low enrichment uranium (LEU) fuel. Different conceptual designs are being developed based on the facility mission and the engineering requirements including nuclear physics, neutronics, heat transfer, thermal hydraulics, structure, and material issues. Different fuel designs with LEU and reflector materials are considered in the design process. Safety, reliability, and environmental considerations are included in the facility conceptual design. The facility is configured to accommodate future design improvements and upgrades. This report is a part of the Argonne National Laboratory Activity within this collaboration for developing and characterizing the subcritical assembly conceptual design. In this study, the medical isotope production function of the Kharkov facility is defined. First, a review was carried out to identify the medical isotopes and its medical use. Then a preliminary assessment was performed without including the self-shielding effect of the irradiated samples. Finally, more detailed investigation was carried out including the self-shielding effect, which defined the sample size and irradiation location for producing each medical isotope. In the first part, the reaction rates were calculated as the multiplication of the cross section with the unperturbed neutron flux of the facility. Over fifty isotopes were considered and all transmutation channels are used including (n,{gamma}), (n,2n), (n,p), and ({gamma},n). In the second part, the parent isotopes with high reaction rate were explicitly modeled in the calculations. For the nuclides with a very high capture microscopic cross section, such as iridium, rhenium, and samarium, their specific activities are reduced by a factor of 30 when the self-shielding effect is included. Four irradiation locations were considered in the analyses to maximize the medical isotope production rate. The results show the self-shield effect reduces the specific activity values and changes the irradiation location for obtaining the maximum possible specific activity. The axial and radial distributions of the specific activity were used to define the irradiation sample size for producing each isotope.

Talamo, A.; Gohar, Y.; Nuclear Engineering Division

2007-05-15T23:59:59.000Z

267

Isotopic Generation and Confirmation of the PWR Application Model   

SciTech Connect

The objective of this calculation is to establish an isotopic database to represent commercial spent nuclear fuel (CSNF) from pressurized water reactors (PWRs) in criticality analyses performed for the proposed Monitored Geologic Repository at Yucca Mountain, Nevada. Confirmation of the conservatism with respect to criticality in the isotopic concentration values represented by this isotopic database is performed as described in Section 3.5.3.1.2 of the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000). The isotopic database consists of the set of 14 actinides and 15 fission products presented in Section 3.5.2.1.1 of YMP 2000 for use in CSNF burnup credit. This set of 29 isotopes is referred to as the principal isotopes. The oxygen isotope from the UO{sub 2} fuel is also included in the database. The isotopic database covers enrichments of {sup 235}U ranging from 1.5 to 5.5 weight percent (wt%) and burnups ranging from approximately zero to 75 GWd per metric ton of uranium (mtU). The choice of fuel assembly and operating history values used in generating the isotopic database are provided is Section 5. Tables of isotopic concentrations for the 29 principal isotopes (plus oxygen) as a function of enrichment and burnup are provided in Section 6.1. Results of the confirmation of the conservatism with respect to criticality in the isotopic concentration values are provided in Section 6.2.

L.B. Wimmer

2003-11-10T23:59:59.000Z

268

Glossary Term - Isotope  

NLE Websites -- All DOE Office Websites (Extended Search)

Helios Previous Term (Helios) Glossary Main Index Next Term (Joule) Joule Isotope The Three Isotopes of Hydrogen - Protium, Deuterium and Tritium Atoms that have the same number of...

269

Enforcement Letter, International Isotopes Idaho Inc - August 20, 1999 |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

International Isotopes Idaho Inc - August 20, International Isotopes Idaho Inc - August 20, 1999 Enforcement Letter, International Isotopes Idaho Inc - August 20, 1999 August, 20, 1999 Issued to International Isotopes Idaho, Inc. related to the Relocation of an Irradiated Pellet at the Test Reactor Area Hot Cell Facility at the Idaho National Engineering and Environmental Laboratory This letter refers to the Department of Energy's (DOE) evaluation of the facts and circumstances concerning the relocation of an irradiated [isotope] pellet from within a hot cell to an adjoining, outside, charging port service area. This incident occurred on January 6, 1999, at the Idaho National Engineering and Environmental Laboratory's Test Reactor Area Hot Cell Facility (TRA-632). Building TRA-632 is utilized by International

270

Enforcement Letter, International Isotopes Idaho Inc - August 20, 1999 |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

International Isotopes Idaho Inc - August 20, International Isotopes Idaho Inc - August 20, 1999 Enforcement Letter, International Isotopes Idaho Inc - August 20, 1999 August, 20, 1999 Issued to International Isotopes Idaho, Inc. related to the Relocation of an Irradiated Pellet at the Test Reactor Area Hot Cell Facility at the Idaho National Engineering and Environmental Laboratory This letter refers to the Department of Energy's (DOE) evaluation of the facts and circumstances concerning the relocation of an irradiated [isotope] pellet from within a hot cell to an adjoining, outside, charging port service area. This incident occurred on January 6, 1999, at the Idaho National Engineering and Environmental Laboratory's Test Reactor Area Hot Cell Facility (TRA-632). Building TRA-632 is utilized by International

271

Nuclear Reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactors Nuclear reactors created not only large amounts of plutonium needed for the weapons programs, but a variety of other interesting and useful radioisotopes. They produced...

272

Stable isotope studies  

SciTech Connect

The research has been in four general areas: (1) correlation of isotope effects with molecular forces and molecular structures, (2) correlation of zero-point energy and its isotope effects with molecular structure and molecular forces, (3) vapor pressure isotope effects, and (4) fractionation of stable isotopes. 73 refs, 38 figs, 29 tabs.

Ishida, T.

1992-01-01T23:59:59.000Z

273

Method for separating isotopes  

DOE Patents (OSTI)

Isotopes are separated by contacting a feed solution containing the isotopes with a cyclic polyether wherein a complex of one isotope is formed with the cyclic polyether, the cyclic polyether complex is extracted from the feed solution, and the isotope is thereafter separated from the cyclic polyether.

Jepson, B.E.

1975-10-21T23:59:59.000Z

274

CRAD, Configuration Management - Oak Ridge National Laboratory...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Configuration Management - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Configuration Management - Oak Ridge National Laboratory High Flux Isotope...

275

Isotopes: Isotope Production, Medical IsotopesOffice of Science...  

NLE Websites -- All DOE Office Websites (Extended Search)

Managers Put a short description of the graphic or its primary message here Isotope Production and Applications The Los Alamos National Laboratory has produced radioactive...

276

Isotope separation by photochromatography  

DOE Patents (OSTI)

An isotope separation method which comprises physically adsorbing an isotopically mixed molecular species on an adsorptive surface and irradiating the adsorbed molecules with radiation of a predetermined wavelength which will selectively excite a desired isotopic species. Sufficient energy is transferred to the excited molecules to desorb them from the surface and thereby separate them from the unexcited undesired isotopic species. The method is particularly applicable to the separation of hydrogen isotopes.

Suslick, Kenneth S. (Stanford, CA)

1977-01-01T23:59:59.000Z

277

Reactor monitoring and safeguards using antineutrino detectors  

E-Print Network (OSTI)

Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactors, as part of International Atomic Energy Agency (IAEA) and other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway across the globe.

N. S. Bowden

2008-09-12T23:59:59.000Z

278

Reactor monitoring and safeguards using antineutrino detectors  

E-Print Network (OSTI)

Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactors, as part of International Atomic Energy Agency (IAEA) and other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway across the globe.

Bowden, N S

2008-01-01T23:59:59.000Z

279

CONTROL SYSTEM FOR NEUTRONIC REACTORS  

DOE Patents (OSTI)

BS>A slow-acting shim rod for control of major variations in reactor neutron flux and a fast-acting control rod to correct minor flux variations are employed to provide a sensitive, accurate control system. The fast-acting rod is responsive to an error signal which is produced by changes in the neutron flux from a predetermined optimum level. When the fast rod is thus actuated in a given direction, means is provided to actuate the slow-moving rod in that direction to return the fast rod to a position near the midpoint of its control range. (AEC)

Crever, F.E.

1962-05-01T23:59:59.000Z

280

MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

Balent, R.

1963-03-12T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

Nuclear reactor shield including magnesium oxide  

DOE Patents (OSTI)

An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.

Rouse, Carl A. (Del Mar, CA); Simnad, Massoud T. (La Jolla, CA)

1981-01-01T23:59:59.000Z

282

Laser-assisted isotope separation of tritium  

DOE Patents (OSTI)

Methods for laser-assisted isotope separation of tritium, using infrared multiple photon dissociation of tritium-bearing products in the gas phase. One such process involves the steps of (1) catalytic exchange of a deuterium-bearing molecule XYD with tritiated water DTO from sources such as a heavy water fission reactor, to produce the tritium-bearing working molecules XYT and (2) photoselective dissociation of XYT to form a tritium-rich product. By an analogous procedure, tritium is separated from tritium-bearing materials that contain predominately hydrogen such as a light water coolant from fission or fusion reactors.

Herman, Irving P. (Castro Valley, CA); Marling, Jack B. (Livermore, CA)

1983-01-01T23:59:59.000Z

283

Glossary of Terms (1mb)  

Science Conference Proceedings (OSTI)

High Flux Isotope Reactor (HFIR) Oak Ridge National Laboratory. Operating at 85 MW, HFIR is the highest flux reactor-based source of neutrons for research in  ...

284

Geometrical vector flux sinks and ideal flux concentrators  

SciTech Connect

The description of ideal flux concentrators as shapes that do not disturb the geometrical vector flux field is extended to all the known types of ideal flux concentrators. This is accomplished, in part, by the introduction of vector flux sinks.

Greenman, P.

1981-06-01T23:59:59.000Z

285

Overview of the US-Japan collaborative investigation on hydrogen isotope retention in neutron-irradiated and ion-damaged tungsten  

Science Conference Proceedings (OSTI)

Plasma-facing components (PFCs) will be exposed to 14 MeV neutrons from deuterium-tritium (D-T) fusion reactions, and tungsten, a candidate PFC for the divertor in ITER, is expected to receive a neutron dose of 0.7 displacement per atom (dpa) by the end of operation in ITER. The effect of neutron-irradiation damage has been mainly simulated using high-energy ion bombardment. While this prior database of results is quite valuable for understanding the behavior of hydrogen isotopes in PFCs, it does not encompass the full range of effects that must be considered in a practical fusion environment due to short penetration depth, damage gradient, high damage rate, and high PKA energy spectrum of the ion bombardment. In addition, neutrons change the elemental composition via transmutations, and create a high radiation environment inside PFCs, which influence the behavior of hydrogen isotope in PFCs, suggesting the utilization of fission reactors is necessary for neutron irradiation. Therefore, the effort to correlate among high-energy ions, fission neutrons, and fusion neutrons is crucial for accurately estimating tritium retention under a neutron-irradiation environment. Under the framework of the US-Japan TITAN program, tungsten samples (99.99 at. % purity from A.L.M.T. Co.) were irradiated by neutron in the High Flux Isotope Reactor (HFIR), ORNL, at 50 and 300C to 0.025, 0.3, and 1.2 dpa, and the investigation of deuterium retention in neutron-irradiation was performed in the INL Tritium Plasma Experiment (TPE), the unique high-flux linear plasma facility that can handle tritium, beryllium and activated materials. This paper reports the recent results from the comparison of ion-damaged tungsten via various ion species (2.8 MeV Fe2+, 20 MeV W2+, and 700 keV H-) with that from neutron-irradiated tungsten to identify the similarities and differences among them.

Masashi Shimada; Y. Hatano; Y. Oya; T. Oda; M. Hara; G. Cao; M. Kobayashi; M. Sokolov; H. Watanabe; B. Tyburska; Y. Ueda; P. Calderoni

2011-09-01T23:59:59.000Z

286

NUCLEAR REACTOR  

DOE Patents (OSTI)

A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

Treshow, M.

1961-09-01T23:59:59.000Z

287

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

Daniels, F.

1959-10-27T23:59:59.000Z

288

Isotope Enrichment Calculator  

Science Conference Proceedings (OSTI)

... incremental isotopic percentages which are compared with an input experimentally derived profile. The theoretical profile of 15 N percentage which ...

2012-10-09T23:59:59.000Z

289

CONVECTION REACTOR  

DOE Patents (OSTI)

An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

Hammond, R.P.; King, L.D.P.

1960-03-22T23:59:59.000Z

290

Expert Panel: Forecast Future Demand for Medical Isotopes | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Expert Panel: Forecast Future Demand for Medical Isotopes Expert Panel: Forecast Future Demand for Medical Isotopes Expert Panel: Forecast Future Demand for Medical Isotopes The Expert Panel has concluded that the Department of Energy and National Institutes of Health must develop the capability to produce a diverse supply of radioisotopes for medical use in quantities sufficient to support research and clinical activities. Such a capability would prevent shortages of isotopes, reduce American dependence on foreign radionuclide sources and stimulate biomedical research. The expert panel recommends that the U.S. government build this capability around either a reactor, an accelerator or a combination of both technologies as long as isotopes for clinical and research applications can be supplied reliably, with diversity in adequate

291

Oceanic Heat Flux Calculation  

Science Conference Proceedings (OSTI)

The authors review the procedure for the direct calculation of oceanic heat flux from hydrographic measurements and set out the full “recipe” that is required.

Sheldon Bacon; Nick Fofonoff

1996-12-01T23:59:59.000Z

292

Isotopically controlled semiconductors  

SciTech Connect

Semiconductor bulk crystals and multilayer structures with controlled isotopic composition have attracted much scientific and technical interest in the past few years. Isotopic composition affects a large number of physical properties, including phonon energies and lifetimes, bandgaps, the thermal conductivity and expansion coefficient and spin-related effects. Isotope superlattices are ideal media for self-diffusion studies. In combination with neutron transmutation doping, isotope control offers a novel approach to metal-insulator transition studies. Spintronics, quantum computing and nanoparticle science are emerging fields using isotope control.

Haller, Eugene E.

2001-12-21T23:59:59.000Z

293

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

SciTech Connect

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.

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

1983-03-01T23:59:59.000Z

294

Advances in reactor physics education: Visualization of reactor parameters  

Science Conference Proceedings (OSTI)

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

Snoj, L.; Kromar, M.; Zerovnik, G. [Josef Stefan Inst., Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

2012-07-01T23:59:59.000Z

295

Candidate processes for diluting the {sup 235}U isotope in weapons-capable highly enriched uranium  

SciTech Connect

The United States Department of Energy (DOE) is evaluating options for rendering its surplus inventories of highly enriched uranium (HEU) incapable of being used to produce nuclear weapons. Weapons-capable HEU was earlier produced by enriching uranium in the fissile {sup 235}U isotope from its natural occurring 0.71 percent isotopic concentration to at least 20 percent isotopic concentration. Now, by diluting its concentration of the fissile {sup 235}U isotope in a uranium blending process, the weapons capability of HEU can be eliminated in a manner that is reversible only through isotope enrichment, and therefore, highly resistant to proliferation. To the extent that can be economically and technically justified, the down-blended uranium product will be made suitable for use as commercial reactor fuel. Such down-blended uranium product can also be disposed of as waste if chemical or isotopic impurities preclude its use as reactor fuel.

Snider, J.D.

1996-02-01T23:59:59.000Z

296

ARM - Measurement - Isotope ratio  

NLE Websites -- All DOE Office Websites (Extended Search)

govMeasurementsIsotope ratio govMeasurementsIsotope ratio ARM Data Discovery Browse Data Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Isotope ratio Ratio of stable isotope concentrations. Categories Atmospheric Carbon, Atmospheric State Instruments The above measurement is considered scientifically relevant for the following instruments. Refer to the datastream (netcdf) file headers of each instrument for a list of all available measurements, including those recorded for diagnostic or quality assurance purposes. ARM Instruments FLASK : Flask Samplers for Carbon Cycle Gases and Isotopes Field Campaign Instruments FLASK : Flask Samplers for Carbon Cycle Gases and Isotopes Datastreams FLASK : Flask Samplers for Carbon Cycle Gases and Isotopes

297

Advanced Test Reactor National Scientific User Facility Partnerships  

SciTech Connect

In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin-Madison; (8) Illinois Institute of Technology (IIT) Materials Research Collaborative Access Team (MRCAT) beamline at Argonne National Laboratory's Advanced Photon Source; and (9) Nanoindenter in the University of California at Berkeley (UCB) Nuclear Engineering laboratory Materials have been analyzed for ATR NSUF users at the Advanced Photon Source at the MRCAT beam, the NIST Center for Neutron Research in Gaithersburg, MD, the Los Alamos Neutron Science Center, and the SHaRE user facility at Oak Ridge National Laboratory (ORNL). Additionally, ORNL has been accepted as a partner facility to enable ATR NSUF users to access the facilities at the High Flux Isotope Reactor and related facilities.

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

2012-03-01T23:59:59.000Z

298

NUCLEAR TRAINING REACTOR. Preliminary Report and Hazards Analysis  

SciTech Connect

Complete descriptions of the reactor, building, and site are given. Reactor operation, bazards associated with the reactor, and the maximum credible accident are discussed. The reactor is a BSR swimming pool type with a design power level of 10 kw and a maximum thermal flux in the fuel region of 2.23 x 10/ sup 11/ n/cm/sup 2/-sec. (W.D.M.)

1959-09-01T23:59:59.000Z

299

FUEL ASSAY REACTOR  

DOE Patents (OSTI)

A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

1962-12-25T23:59:59.000Z

300

WATER BOILER REACTOR  

DOE Patents (OSTI)

As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

King, L.D.P.

1960-11-22T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Concentration with uniform flux  

SciTech Connect

A modification of a parabolic cylinder concentrator is developed to procedure uniform flux. The controlling surface equation is given. A three-dimensional ray-trace technique is used to obtain the shape of the image at the focal plane of a thin slice of the mirror. Also, the concentration distribution for uniform flux is given. 1 references, 7 figures.

Not Available

1986-01-01T23:59:59.000Z

302

Nuclear Reactor Safeguards and Monitoring with Antineutrino Detectors  

E-Print Network (OSTI)

Cubic-meter-sized antineutrino detectors can be used to non-intrusively, robustly and automatically monitor and safeguard a wide variety of nuclear reactor types, including power reactors, research reactors, and plutonium production reactors. Since the antineutrino spectra and relative yields of fissioning isotopes depend on the isotopic composition of the core, changes in composition can be observed without ever directly accessing the core itself. Information from a modest-sized antineutrino detector, coupled with the well-understood principles that govern the core's evolution in time, can be used to determine whether the reactor is being operated in an illegitimate way. A group at Sandia is currently constructing a one cubic meter antineutrino detector at the San Onofre reactor site in California to demonstrate these principles.

Adam Bernstein; Yifang Wang; Giorgio Gratta; Todd West

2001-08-01T23:59:59.000Z

303

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

Fraas, A.P.; Mills, C.B.

1961-11-21T23:59:59.000Z

304

REACTOR COOLING  

DOE Patents (OSTI)

A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

Quackenbush, C.F.

1959-09-29T23:59:59.000Z

305

CONTROL MEANS FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

A control means is described for a reactor which employs a liquid fuel consisting of a fissile isotope in a liquid bismuth solvent. The liquid fuel is contained in a plurality of tubular vessels. Control is effected by inserting plungers in the vessels to displace the liquid fuel and provide a critical or non- critical fuel configuration as desired.

Teitel, R.J.

1961-09-01T23:59:59.000Z

306

Uranium mill monitoring for natural fission reactors  

SciTech Connect

Isotopic monitoring of the product stream from operating uranium mills is proposed for discovering other possible natural fission reactors; aspects of their occurrence and discovery are considered. Uranium mill operating characteristics are formulated in terms of the total uranium capacity, the uranium throughput, and the dilution half-time of the mill. The requirements for detection of milled reactor-zone uranium are expressed in terms of the dilution half-time and the sampling frequency. Detection of different amounts of reactor ore with varying degrees of /sup 235/U depletion is considered.

Apt, K.E.

1977-12-01T23:59:59.000Z

307

Atomic vapor laser isotope separation  

SciTech Connect

Atomic vapor laser isotope separation (AVLIS) is a general and powerful technique. A major present application to the enrichment of uranium for light-water power reactor fuel has been under development for over 10 years. In June 1985 the Department of Energy announced the selection of AVLIS as the technology to meet the nation's future need for the internationally competitive production of uranium separative work. The economic basis for this decision is considered, with an indicated of the constraints placed on the process figures of merit and the process laser system. We then trace an atom through a generic AVLIS separator and give examples of the physical steps encountered, the models used to describe the process physics, the fundamental parameters involved, and the role of diagnostic laser measurements.

Stern, R.C.; Paisner, J.A.

1985-11-08T23:59:59.000Z

308

Hybrid isotope separation scheme  

DOE Patents (OSTI)

A method of yielding selectively a desired enrichment in a specific isotope including the steps of inputting into a spinning chamber a gas from which a scavenger, radiating the gas with a wave length or frequency characteristic of the absorption of a particular isotope of the atomic or molecular gas, thereby inducing a photochemical reaction between the scavenger, and collecting the specific isotope-containing chemical by using a recombination surface or by a scooping apparatus.

Maya, Jakob (Brookline, MA)

1991-01-01T23:59:59.000Z

309

HYDROGEN ISOTOPE TARGETS  

DOE Patents (OSTI)

The design of targets for use in the investigation of nuclear reactions of hydrogen isotopes by bombardment with accelerated particles is described. The target con struction eomprises a backing disc of a metal selected from the group consisting of molybdenunn and tungsten, a eoating of condensed titaniunn on the dise, and a hydrogen isotope selected from the group consisting of deuterium and tritium absorbed in the coatiag. The proeess for preparing these hydrogen isotope targets is described.

Ashley, R.W.

1958-08-12T23:59:59.000Z

310

Computing Solar Absolute Fluxes  

E-Print Network (OSTI)

Computed color indices and spectral shapes for individual stars are routinely compared with observations for essentially all spectral types, but absolute fluxes are rarely tested. We can confront observed irradiances with the predictions from model atmospheres for a few stars with accurate angular diameter measurements, notably the Sun. Previous calculations have been hampered by inconsistencies and the use of outdated atomic data and abundances. I provide here a progress report on our current efforts to compute absolute fluxes for solar model photospheres. Uncertainties in the solar composition constitute a significant source of error in computing solar radiative fluxes.

Prieto, Carlos Allende

2007-01-01T23:59:59.000Z

311

Computing Solar Absolute Fluxes  

E-Print Network (OSTI)

Computed color indices and spectral shapes for individual stars are routinely compared with observations for essentially all spectral types, but absolute fluxes are rarely tested. We can confront observed irradiances with the predictions from model atmospheres for a few stars with accurate angular diameter measurements, notably the Sun. Previous calculations have been hampered by inconsistencies and the use of outdated atomic data and abundances. I provide here a progress report on our current efforts to compute absolute fluxes for solar model photospheres. Uncertainties in the solar composition constitute a significant source of error in computing solar radiative fluxes.

Carlos Allende Prieto

2007-09-14T23:59:59.000Z

312

Detection of the Neutrino Fluxes from Several Sources  

E-Print Network (OSTI)

It is considered the detection of neutrinos moving from the opposite directions. The states of the particle of the detector interacting with the neutrinos are connected with the P-transformation. Hence only a half of neutrinos gives contribution into the superposition of the neutrino states. Taking into account the effect of the opposite neutrino directions the total neutrino flux from several sources are in the range 0.5--1 of that without the effect. The neutrino flux from nuclear reactors measured in the KamLAND experiment is $0.611\\pm 0.085 {\\rm (stat)} \\pm 0.041 {\\rm (syst)} $ of the expected flux. Calculations for the conditions of the KamLAND experiment yield the neutrino flux taking into account the effect of the opposite neutrino directions, 0.555, of that without the effect that may account for the neutrino flux observed in the KamLAND experiment.

D. L. Khokhlov

2003-02-19T23:59:59.000Z

313

Isotopic Bias and Uncertainty for Burnup Credit Applications  

Science Conference Proceedings (OSTI)

The application of burnup credit requires calculating the isotopic inventory of the irradiated fuel. The depletion calculation simulates the burnup of the fuel under reactor operating conditions. The result of the depletion analysis is the predicted isotopic composition, which is ultimately input to a criticality analysis to determine the system multiplication factor (k{sub eff}). This paper demonstrates an approach for calculating the isotopic bias and uncertainty in k{sub eff} for commercial spent nuclear fuel burnup credit. This paper covers 74 different radiochemical assayed spent fuel samples from 22 different fuel assemblies that were irradiated in eight different pressurized water reactors (PWRs). The samples evaluated span an enrichment range of 2.556 wt% U-235 through 4.67 wt% U-235, and burnups from 6.92 GWd/MTU through 55.7 GWd/MTU.

J.M. Scaglione

2002-08-19T23:59:59.000Z

314

Photovoltaic roof heat flux  

E-Print Network (OSTI)

of ~24°C, indicating that heat conduction was small. T h i sday, indicating large heat conduction a n d storage. Control2.1.3 showed that conduction heat flux through the roof was

Samady, Mezhgan Frishta

2011-01-01T23:59:59.000Z

315

Discovery of the Mercury Isotopes  

E-Print Network (OSTI)

Forty mercury isotopes have so far been observed; the discovery of these isotopes is discussed. For each isotope a brief summary of the first refereed publication, including the production and identification method, is presented.

D. Meierfrankenfeld; M. Thoennessen

2009-12-01T23:59:59.000Z

316

The nucifer experiment: Non-proliferation with reactor antineutrinos  

Science Conference Proceedings (OSTI)

A survey of the reactor antineutrino flux provides information related to the core content and thermal power. This application arouses the interest of International Atomic Energy Agency (IAEA) in using antineutrino detectors as a potential safeguard tool. Within this context

A. S. Cucoanes; Nucifer Collaboration

2012-01-01T23:59:59.000Z

317

Advanced fuel fusion reactors: towards a zero-waste option  

E-Print Network (OSTI)

Low activation materials are only a partial response to the requirement of a really environmentally sound fusion reactor: another way round to tackle the problem is the reduction of the neutron flux and subsequent material ...

Zucchetti, Massimo

318

Thermionic switched self-actuating reactor shutdown system  

DOE Patents (OSTI)

A self-actuating reactor shutdown system incorporating a thermionic switched electromagnetic latch arrangement which is responsive to reactor neutron flux changes and to reactor coolant temperature changes. The system is self-actuating in that the sensing thermionic device acts directly to release (scram) the control rod (absorber) without reference or signal from the main reactor plant protective and control systems. To be responsive to both temperature and neutron flux effects, two detectors are used, one responsive to reactor coolant temperatures, and the other responsive to reactor neutron flux increase. The detectors are incorporated into a thermionic diode connected electrically with an electromagnetic mechanism which under normal reactor operating conditions holds the the control rod in its ready position (exterior of the reactor core). Upon reaching either a specified temperature or neutron flux, the thermionic diode functions to short-circuit the electromagnetic mechanism causing same to lose its holding power and release the control rod, which drops into the reactor core region under gravitational force.

Barrus, Donald M. (San Jose, CA); Shires, Charles D. (San Jose, CA); Brummond, William A. (Livermore, CA)

1989-01-01T23:59:59.000Z

319

Direct evaluation of transient surface temperatures and heat fluxes  

SciTech Connect

Evaluations of transient surface temperatures resulting from the absorption of radiation are required in laser fusion reactor systems studies. A general method for the direct evaluation of transient surface temperatures and heat fluxes on the boundaries of bounded media is developed by constructing fundamental solutions of the scalar Helmholtz equation and performing certain elementary integrations. (auth)

Axford, R.A.

1975-08-01T23:59:59.000Z

320

Level 1 transient model for a molybdenum-99 producing aqueous homogeneous reactor and its applicability to the tracy reactor  

SciTech Connect

Babcock and Wilcox Technical Services Group (B and W) has identified aqueous homogeneous reactors (AHRs) as a technology well suited to produce the medical isotope molybdenum 99 (Mo-99). AHRs have never been specifically designed or built for this specialized purpose. However, AHRs have a proven history of being safe research reactors. In fact, in 1958, AHRs had 'a longer history of operation than any other type of research reactor using enriched fuel' and had 'experimentally demonstrated to be among the safest of all various type of research reactor now in use [1].' A 'Level 1' model representing B and W's proposed Medical Isotope Production System (MIPS) reactor has been developed. The Level 1 model couples a series of differential equations representing neutronics, temperature, and voiding. Neutronics are represented by point reactor kinetics while temperature and voiding terms are axially varying (one-dimensional). While this model was developed specifically for the MIPS reactor, its applicability to the Japanese TRACY reactor was assessed. The results from the Level 1 model were in good agreement with TRACY experimental data and found to be conservative over most of the time domains considered. The Level 1 model was used to study the MIPS reactor. An analysis showed the Level 1 model agreed well with a more complex computational model of the MIPS reactor (a FETCH model). Finally, a significant reactivity insertion was simulated with the Level 1 model to study the MIPS reactor's time-dependent response. (authors)

Nygaard, E. T. [Babcock and Wilcox Technical Services Group, 800 Main Street, Lynchburg, VA 24504 (United States); Williams, M. M. R. [Imperial College London, SW7 2AZ (United Kingdom); Angelo, P. L. [Y-12 National Security Complex, Oak Ridge, TN 37831 (United States)

2012-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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321

The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor  

SciTech Connect

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In addition, the purpose and differences between the two experiments will be compared and the irradiation results to date on the first experiment will be presented.

S. Blaine Grover

2009-09-01T23:59:59.000Z

322

NUCLEAR REACTOR  

DOE Patents (OSTI)

A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

Moore, R.V.; Bowen, J.H.; Dent, K.H.

1958-12-01T23:59:59.000Z

323

Fuel Cycle and Isotopes Division  

NLE Websites -- All DOE Office Websites (Extended Search)

Divisions Fuel Cycle and Isotopes Division Jeffrey Binder, Division Director Jeffrey Binder, Division Director The Fuel Cycle and Isotopes Division (FCID) of the Nuclear Science...

324

Isotopic Analysis | Open Energy Information  

Open Energy Info (EERE)

Structural: Hydrological: Source of fluids, circulation, andor mixing. Thermal: Heat source and general reservoir temperatures Dictionary.png Isotopic Analysis: Isotopes...

325

Isotope Enrichment | ornl.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

Modern electromagnetic isotope separator developed and being scaled-up to replace the Manhattan Project-era Calutrons used for stable isotope enrichment. Since 1945, ORNL has...

326

Laser isotope separation  

DOE Patents (OSTI)

A process and apparatus for separating isotopes by selective excitation of isotopic species of a volatile compound by tuned laser light is described. A highly cooled gas of the volatile compound is produced in which the isotopic shift is sharpened and defined. Before substantial condensation occurs, the cooled gas is irradiated with laser light precisely tuned to a desired wavelength to selectively excite a particular isotopic species in the cooled gas. The laser light may impart sufficient energy to the excited species to cause it to undergo photolysis, photochemical reaction or even to photoionize. Alternatively, a two-photon irradiation may be applied to the cooled gas to induce photolysis, photochemical reaction or photoionization. The process is particularly applicable to the separation of isotopes of uranium.

Robinson, C.P.; Reed, J.J.; Cotter, T.P.; Boyer, K.; Greiner, N.R.

1975-11-26T23:59:59.000Z

327

Isotope GeochemistryIsotope Geochemistry Isotopes do not fractionate during partial  

E-Print Network (OSTI)

/204Pb, 207Pb/204Pb, due to U and Th decay The isotope geology of PbThe isotope geology of Pb #12;The isotope geology of PbThe isotope geology of Pb µ = 238U/204Pb Primeval lead (Isotope ratios of Pb tT t eea Pb Pb -µ+= 30.90 204 206 == a Pb Pb i 29.100 204 207 == b Pb Pb i #12;The isotope geology

Siebel, Wolfgang

328

Reactor Materials  

Energy.gov (U.S. Department of Energy (DOE))

The reactor materials crosscut effort will enable the development of innovative and revolutionary materials and provide broad-based, modern materials science that will benefit all four DOE-NE...

329

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

1962-10-23T23:59:59.000Z

330

NEUTRONIC REACTORS  

DOE Patents (OSTI)

A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

Wigner, E.P.

1960-11-22T23:59:59.000Z

331

REACTOR SHIELD  

DOE Patents (OSTI)

Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

1959-02-17T23:59:59.000Z

332

Flux lattices reformulated  

E-Print Network (OSTI)

We theoretically explore the optical flux lattices produced for ultra-cold atoms subject to laser fields where both the atom-light coupling and the effective detuning are spatially periodic. We analyze the geometric vector potential and the magnetic flux it generates, as well as the accompanying geometric scalar potential. We show how to understand the gauge-dependent Aharonov-Bohm singularities in the vector potential, and calculate the continuous magnetic flux through the elementary cell in terms of these singularities. The analysis is illustrated with a square optical flux lattice. We conclude with an explicit laser configuration yielding such a lattice using a set of five properly chosen beams with two counterpropagating pairs (one along the x axes and the other y axes), together with a single beam along the z axis. We show that this lattice is not phase-stable, and identify the one phase-difference that affects the magnetic flux. Thus armed with realistic laser setup, we directly compute the Chern number...

Juzeli?nas, G

2012-01-01T23:59:59.000Z

333

NUCLEAR REACTOR  

DOE Patents (OSTI)

High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

Grebe, J.J.

1959-07-14T23:59:59.000Z

334

Advanced Safeguards Approaches for New Fast Reactors  

SciTech Connect

This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

2007-12-15T23:59:59.000Z

335

On fast reactor kinetics studies  

Science Conference Proceedings (OSTI)

The results and the program of fast reactor core time and space kinetics experiments performed and planned to be performed at the IPPE critical facility is presented. The TIMER code was taken as computation support of the experimental work, which allows transient equations to be solved in 3-D geometry with multi-group diffusion approximation. The number of delayed neutron groups varies from 6 to 8. The code implements the solution of both transient neutron transfer problems: a direct one, where neutron flux density and its derivatives, such as reactor power, etc, are determined at each time step, and an inverse one for the point kinetics equation form, where such a parameter as reactivity is determined with a well-known reactor power time variation function. (authors)

Seleznev, E. F.; Belov, A. A. [Nuclear Safety Inst. of the Russian Academy of Sciences IBRAE (Russian Federation); Matveenko, I. P.; Zhukov, A. M.; Raskach, K. F. [Inst. for Physics and Power Engineering IPPE (Russian Federation)

2012-07-01T23:59:59.000Z

336

Solid State Reactor Final Report  

DOE Green Energy (OSTI)

The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas of research were undertaken: (1) establishing the design and safety-related basis via neutronic and reactor control assessments with the graphite foam as heat transfer medium; (2) evaluating the thermal performance of the graphite foam for heat removal, reactor stability, reactor operations, and overall core thermal characteristics; (3) characterizing the physical properties of the graphite foam under normal and irradiated conditions to determine any effects on structure, dimensional stability, thermal conductivity, and thermal expansion; and (4) developing a power conversion system design to match the reactor operating parameters.

Mays, G.T.

2004-03-10T23:59:59.000Z

337

Optical heat flux gauge  

DOE Patents (OSTI)

A heat flux gauge comprising first and second thermographic phosphor layers separated by a layer of a thermal insulator. The gauge may be mounted on a surface with the first thermographic phosphor in contact with the surface. A light source is directed at the gauge, causing the phosphors to luminesce. The luminescence produced by the phosphors is collected and its spectra analyzed in order to determine the heat flux on the surface. First and second phosphor layers must be different materials to assure that the spectral lines collected will be distinguishable. 9 figs.

Noel, B.W.; Borella, H.M.; Cates, M.R.; Turley, W.D.; MacArthur, C.D.; Cala, G.C.

1989-06-07T23:59:59.000Z

338

Isotopically controlled semiconductors  

SciTech Connect

The following article is an edited transcript based on the Turnbull Lecture given by Eugene E. Haller at the 2005 Materials Research Society Fall Meeting in Boston on November 29, 2005. The David Turnbull Lectureship is awarded to recognize the career of a scientist who has made outstanding contributions to understanding materials phenomena and properties through research, writing, and lecturing, as exemplified by the life work of David Turnbull. Haller was named the 2005 David Turnbull Lecturer for his 'pioneering achievements and leadership in establishing the field of isotopically engineered semiconductors; for outstanding contributions to materials growth, doping and diffusion; and for excellence in lecturing, writing, and fostering international collaborations'. The scientific interest, increased availability, and technological promise of highly enriched isotopes have led to a sharp rise in the number of experimental and theoretical studies with isotopically controlled semiconductor crystals. This article reviews results obtained with isotopically controlled semiconductor bulk and thin-film heterostructures. Isotopic composition affects several properties such as phonon energies, band structure, and lattice constant in subtle, but, for their physical understanding, significant ways. Large isotope-related effects are observed for thermal conductivity in local vibrational modes of impurities and after neutron transmutation doping. Spectacularly sharp photoluminescence lines have been observed in ultrapure, isotopically enriched silicon crystals. Isotope multilayer structures are especially well suited for simultaneous self- and dopant-diffusion studies. The absence of any chemical, mechanical, or electrical driving forces makes possible the study of an ideal random-walk problem. Isotopically controlled semiconductors may find applications in quantum computing, nanoscience, and spintronics.

Haller, Eugene E.

2006-06-19T23:59:59.000Z

339

RERTR program reduces use of enriched uranium in research reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

RERTR program reduces use of enriched uranium in research reactors RERTR program reduces use of enriched uranium in research reactors worldwide Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear Reactors Designed by Argonne Argonne's Nuclear Science and Technology Legacy Opportunities within NE Division Visit Argonne Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Celebrating the 70th Anniversary of Chicago Pile 1 (CP-1) Argonne OutLoud on Nuclear Energy Argonne Energy Showcase 2012 Highlights Bookmark and Share RERTR program reduces use of enriched uranium in research reactors worldwide The High Flux Reactor in Petten, the Netherlands READY TO CONVERT - The High Flux Reactor in Petten, the Netherlands, has

340

Brookhaven Medical Research Reactor  

NLE Websites -- All DOE Office Websites (Extended Search)

Medical Research Reactor BMRR The last of the Lab's reactors, the Brookhaven Medical Research Reactor (BMRR), was shut down in December 2000. The BMRR was a three megawatt...

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

NEUTRONIC REACTOR  

DOE Patents (OSTI)

This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

1958-09-01T23:59:59.000Z

342

Mapping Heat Flux  

Science Conference Proceedings (OSTI)

An infrared camera technique designed for remote sensing of air–water heat flux has been developed. The technique uses the differential absorption of water between 3.817 and 4.514 microns. This difference causes each channel’s radiance to ...

Walt McKeown; Richard Leighton

1999-01-01T23:59:59.000Z

343

GALLIUM ARSENIDE SEMICONDUCTOR-BASED NEUTRON DETECTOR  

NEUTRON DETECTOR BENEFITS Portable, ... High Flux Isotope Reactor and Spallation Neutron Source. Several Homeland Security. LINKS TO ONLINE ...

344

Multidisciplinary multi-physics simulation and  

E-Print Network (OSTI)

to accept and process proposals submitted by users for beam time at the High Flux Isotope Reactor (HFIR

345

DOE Designated User Facilities Multiple Laboratories * ARM Climate...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

* Center for Nanophase Materials Sciences (CNMS) * High Flux Isotope Reactor (HFIR) * National Center for Computational Sciences (NCCS) * Shared Research Equipment...

346

Structural Materials Development for MFE and IFE  

E-Print Network (OSTI)

's Spallation Neutron Source/High Flux Isotope Reactor (SNS/HFIR), National Center for Computational Sciences

347

Life Sciences Division Open Literature Publications  

E-Print Network (OSTI)

in the ORNL High Flux Isotope Reactor (HFIR) for Cancer Treatment and Arterial Restenosis Therapy After PTCA

348

Self-actuating reactor shutdown system  

DOE Patents (OSTI)

A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

Barrus, Donald M. (San Jose, CA); Brummond, Willian A (Livermore, CA); Peterson, Leslie F. (Danville, CA)

1988-01-01T23:59:59.000Z

349

REACTOR CONTROL  

DOE Patents (OSTI)

A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

Fortescue, P.; Nicoll, D.

1962-04-24T23:59:59.000Z

350

Catalytic reactor  

DOE Patents (OSTI)

A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

2009-03-10T23:59:59.000Z

351

NEUTRONIC REACTORS  

DOE Patents (OSTI)

A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

Wigner, E.P.; Young, G.J.

1958-10-14T23:59:59.000Z

352

METHOD AND APPARATUS FOR REACTOR SAFETY CONTROL  

DOE Patents (OSTI)

A self-contained nuclear reactor fuse controlled device tron absorbing material, normally in a compact form but which can be expanded into an extended form presenting a large surface for neutron absorption when triggered by an increase in neutron flux, is described.

Huston, N.E.

1961-06-01T23:59:59.000Z

353

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

Young, G.

1963-01-01T23:59:59.000Z

354

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

Wigner, E.P.; Weinberg, A.W.; Young, G.J.

1958-04-15T23:59:59.000Z

355

Power Burst Facility (PBF) Reactor Reactor Decommissioning  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactor Decommissioning Click here to view Click here to view Reactor Decommissioning Click on an image to enlarge A crane removes the reactor vessel from the Power Burst Facility...

356

Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors  

Science Conference Proceedings (OSTI)

Burnable poisons are used in nuclear reactors to produce a more level distribution of power in the reactor core and to reduce to necessity for a large control system. An ideal burnable poison would burn at the same rate as the fuel. In this study, separation of neutron-absorbing isotopes was investigated in order to eliminate isotopes that remain as absorbers at the end of fuel life, thus reducing useful fuel life. The isotopes Gd-157, Dy-164, and Er-167 were found to have desirable properties. These isotopes were separated from naturally occurring elements by means of plasma separation to evaluate feasibility and cost. It was found that pure Gd-157 could save approximately $6 million at the end of four years. However, the cost of separation, using the existing facility, made separation cost- ineffective. Using a magnet with three times the field strength is expected to reduce the cost by a factor of ten, making isotopically separated burnable poisons a favorable method of increasing fuel life in commercial reactors, in particular Generation-IV reactors. The project also investigated various burnable poison configurations, and studied incorporation of metallic burnable poisons into fuel cladding.

M. L. Grossbeck J-P.A. Renier Tim Bigelow

2003-09-30T23:59:59.000Z

357

Welcome - Nuclear Medicine Program  

NLE Websites -- All DOE Office Websites (Extended Search)

combined resources of the stable isotope inventory, a High Flux Isotope Reactor (HFIR), hot cell processing capabilities, and a wide range of support functions required for such...

358

Nuclear Science | More Science | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

of isotopes for medical purposes and research. The lab's High Flux Isotope Reactor (HFIR) and Radiochemical Engineering Development Center (REDC) together provide the western...

359

--No Title--  

NLE Websites -- All DOE Office Websites (Extended Search)

his colleagues are credited with designing the core of the High Flux Isotope Reactor (HFIR). The primary purpose of the HFIR is to produce transplutonium isotopes such as...

360

Deciphering the measured ratios of Iodine-131 to Cesium-137 at the Fukushima reactors  

E-Print Network (OSTI)

We calculate the relative abundance of the radioactive isotopes Iodine-131 and Cesium-137 produced by nuclear fission in reactors and compare it with data taken at the troubled Fukushima Dai-ichi nuclear power plant. The ratio of radioactivities of these two isotopes can be used to obtain information about when the nuclear reactions terminated.

Matsui, T

2011-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Deciphering the measured ratios of Iodine-131 to Cesium-137 at the Fukushima reactors  

E-Print Network (OSTI)

We calculate the relative abundance of the radioactive isotopes Iodine-131 and Cesium-137 produced by nuclear fission in reactors and compare it with data taken at the troubled Fukushima Dai-ichi nuclear power plant. The ratio of radioactivities of these two isotopes can be used to obtain information about when the nuclear reactions terminated.

T. Matsui

2011-05-02T23:59:59.000Z

362

Nuclear Transmutations in HFIR's Beryllium Reflector and Their Impact on Reactor Operation and Reflector Disposal  

SciTech Connect

The High Flux Isotope Reactor located at the Oak Ridge National Laboratory utilizes a large cylindrical beryllium reflector that is subdivided into three concentric regions and encompasses the compact reactor core. Nuclear transmutations caused by neutron activation occur in the beryllium reflector regions, which leads to unwanted neutron absorbing and radiation emitting isotopes. During the past year, two topics related to the HFIR beryllium reflector were reviewed. The first topic included studying the neutron poison (helium-3 and lithium-6) buildup in the reflector regions and its affect on beginning-of-cycle reactivity. A new methodology was developed to predict the reactivity impact and estimated symmetrical critical control element positions as a function of outage time between cycles due to helium-3 buildup and was shown to be in better agreement with actual symmetrical critical control element position data than the current methodology. The second topic included studying the composition of the beryllium reflector regions at discharge as well as during decay to assess the viability of transporting, storing, and ultimately disposing the reflector regions currently stored in the spent fuel pool. The post-irradiation curie inventories were used to determine whether the reflector regions are discharged as transuranic waste or become transuranic waste during the decay period for disposal purposes and to determine the nuclear hazard category, which may affect the controls invoked for transportation and temporary storage. Two of the reflector regions were determined to be transuranic waste at discharge and the other region was determined to become transuranic waste in less than 2 years after being discharged due to the initial uranium content (0.0044 weight percent uranium). It was also concluded that all three of the reflector regions could be classified as nuclear hazard category 3 (potential for localized consequences only).

Chandler, David [ORNL; Maldonado, G Ivan [ORNL; Primm, Trent [ORNL; Proctor, Larry Duane [ORNL

2012-01-01T23:59:59.000Z

363

Isotopes as Environmental Tracers in Archived Biological ...  

Science Conference Proceedings (OSTI)

... Tissue Archival and Monitoring Program (STAMP ... and isotopes) and carbon/nitrogen (isotopes). The carbon/nitrogen isotope data provide valuable ...

2012-10-02T23:59:59.000Z

364

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A reactor is described comprising a plurality of horizontal trays containing a solution of a fissionable material, the trays being sleeved on a vertical tube which contains a vertically-reciprocable control rod, a gas-tight chamber enclosing the trays, and means for conducting vaporized moderator from the chamber and for replacing vaporized moderator in the trays. (AEC)

Wigner, E.P.

1962-12-25T23:59:59.000Z

365

Neutronic reactor  

DOE Patents (OSTI)

A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

Wende, Charles W. J. (West Chester, PA)

1976-08-17T23:59:59.000Z

366

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor is described that includes spaced vertical fuel elements centrally disposed in a pressure vessel, a mass of graphite particles in the pressure vessel, means for fluidizing the graphite particles, and coolant tubes in the pressure vessel laterally spaced from the fuel elements. (AEC)

Post, R.G.

1963-05-01T23:59:59.000Z

367

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

Starr, C.

1963-01-01T23:59:59.000Z

368

NEUTRONIC REACTOR  

DOE Patents (OSTI)

BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

1959-10-27T23:59:59.000Z

369

NEUTRONIC REACTORS  

DOE Patents (OSTI)

The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

Anderson, H.L.

1958-10-01T23:59:59.000Z

370

Hydrogen isotope separation utilizing bulk getters  

DOE Patents (OSTI)

Tritium and deuterium are separated from a gaseous mixture thereof, derived from a nuclear fusion reactor or some other source, by providing a casing with a bulk getter therein for absorbing the gaseous mixture to produce an initial loading of the getter, partially desorbing the getter to produce a desorbed mixture which is tritium-enriched, pumping the desorbed mixture into a separate container, the remaining gaseous loading in the getter being deuterium-enriched, desorbing the getter to a substantially greater extent to produce a deuterium-enriched gaseous mixture, and removing the deuterium-enriched mixture into another container. The bulk getter may comprise a zirconium-aluminum alloy, or a zirconium-vanadium-iron alloy. The partial desorption may reduce the loading by approximately fifty percent. The basic procedure may be extended to produce a multistage isotope separator, including at least one additional bulk getter into which the tritium-enriched mixture is absorbed. The second getter is then partially desorbed to produce a desorbed mixture which is further tritium-enriched. The last-mentioned mixture is then removed from the container for the second getter, which is then desorbed to a substantially greater extent to produce a desorbed mixture which is deuterium-enriched. The last-mentioned mixture is then removed so that the cycle can be continued and repeated. The method of isotope separation is also applicable to other hydrogen isotopes, in that the method can be employed for separating either deuterium or tritium from normal hydrogen.

Knize, Randall J. (Los Angeles, CA); Cecchi, Joseph L. (Lawrenceville, NJ)

1991-01-01T23:59:59.000Z

371

Hydrogen isotope separation utilizing bulk getters  

DOE Patents (OSTI)

Tritium and deuterium are separated from a gaseous mixture thereof, derived from a nuclear fusion reactor or some other source, by providing a casing with a bulk getter therein for absorbing the gaseous mixture to produce an initial loading of the getter, partially desorbing the getter to produce a desorbed mixture which is tritium-enriched, pumping the desorbed mixture into a separate container, the remaining gaseous loading in the getter being deuterium-enriched, desorbing the getter to a substantially greater extent to produce a deuterium-enriched gaseous mixture, and removing the deuterium-enriched mixture into another container. The bulk getter may comprise a zirconium-aluminum alloy, or a zirconium-vanadium-iron alloy. The partial desorption may reduce the loading by approximately fifty percent. The basic procedure may be extended to produce a multistage isotope separator, including at least one additional bulk getter into which the tritium-enriched mixture is absorbed. The second getter is then partially desorbed to produce a desorbed mixture which is further tritium-enriched. The last-mentioned mixture is then removed from the container for the second getter, which is then desorbed to a substantially greater extent to produce a desorbed mixture which is deuterium-enriched. The last-mentioned mixture is then removed so that the cycle can be continued and repeated. The method of isotope separation is also applicable to other hydrogen isotopes, in that the method can be employed for separating either deuterium or tritium from normal hydrogen.

Knize, Randall J. (Los Angeles, CA); Cecchi, Joseph L. (Lawrenceville, NJ)

1990-01-01T23:59:59.000Z

372

Hydrogen isotope separation utilizing bulk getters  

DOE Patents (OSTI)

Tritium and deuterium are separated from a gaseous mixture thereof, derived from a nuclear fusion reactor or some other source, by providing a casing with a bulk getter therein for absorbing the gaseous mixture to produce an initial loading of the getter, partially desorbing the getter to produce a desorbed mixture which is tritium-enriched, pumping the desorbed mixture into a separate container, the remaining gaseous loading in the getter being deuterium-enriched, desorbing the getter to a substantially greater extent to produce a deuterium-enriched gaseous mixture, and removing the deuterium-enriched mixture into another container. The bulk getter may comprise a zirconium-aluminum alloy, or a zirconium-vanadium-iron alloy. The partial desorption may reduce the loading by approximately fifty percent. The basic procedure may be extended to produce a multistage isotope separator, including at least one additional bulk getter into which the tritium-enriched mixture is absorbed. The second getter is then partially desorbed to produce a desorbed mixture which is further tritium-enriched. The last-mentioned mixture is then removed from the container for the second getter, which is then desorbed to a substantially greater extent to produce a desorbed mixture which is deuterium-enriched. The last-mentioned mixture is then removed so that the cycle can be continued and repeated. The method of isotope separation is also applicable to other hydrogen isotopes, in that the method can be employed for separating either deuterium or tritium from normal hydrogen. 4 figures.

Knize, R.J.; Cecchi, J.L.

1991-08-20T23:59:59.000Z

373

Separation of sulfur isotopes  

DOE Patents (OSTI)

Sulfur isotopes are continuously separated and enriched using a closed loop reflux system wherein sulfur dioxide (SO.sub.2) is reacted with sodium hydroxide (NaOH) or the like to form sodium hydrogen sulfite (NaHSO.sub.3). Heavier sulfur isotopes are preferentially attracted to the NaHSO.sub.3, and subsequently reacted with sulfuric acid (H.sub.2 SO.sub.4) forming sodium hydrogen sulfate (NaHSO.sub.4) and SO.sub.2 gas which contains increased concentrations of the heavier sulfur isotopes. This heavy isotope enriched SO.sub.2 gas is subsequently separated and the NaHSO.sub.4 is reacted with NaOH to form sodium sulfate (Na.sub.2 SO.sub.4) which is subsequently decomposed in an electrodialysis unit to form the NaOH and H.sub.2 SO.sub.4 components which are used in the aforesaid reactions thereby effecting sulfur isotope separation and enrichment without objectionable loss of feed materials.

DeWitt, Robert (Centerville, OH); Jepson, Bernhart E. (Dayton, OH); Schwind, Roger A. (Centerville, OH)

1976-06-22T23:59:59.000Z

374

Investigation of downward facing critical heat flux with water-based nanofluids for In-Vessel Retention applications  

E-Print Network (OSTI)

In-Vessel Retention ("IVR") is a severe accident management strategy that is power limiting to the Westinghouse AP1000 due to critical heat flux ("CHF") at the outer surface of the reactor vessel. Increasing the CHF level ...

DeWitt, Gregory L

2011-01-01T23:59:59.000Z

375

Stable Isotopes in Hailstones. Part I: The Isotopic Cloud Model  

Science Conference Proceedings (OSTI)

Equations describing the isotopic balance between five water species (vapor, cloud water, rainwater, cloud ice and graupel)have been incorporated into a one-dimensional steady-state cloud model. The isotope contents of the various water ...

B. Federer; N. Brichet; J. Jouzel

1982-06-01T23:59:59.000Z

376

Preliminary Notice of Violation, Isotopes Idaho, Inc. - EA-2000-04 |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Isotopes Idaho, Inc. - EA-2000-04 Isotopes Idaho, Inc. - EA-2000-04 Preliminary Notice of Violation, Isotopes Idaho, Inc. - EA-2000-04 May 19, 2000 Preliminary Notice of Violation issued to International Isotopes Idaho, Inc., related to Work Planning and Control Deficiencies associated with Replacement of Exhaust Ventilation Filters at the Test Reactor Area Hot Cell Facility at the Idaho National Engineering and Environmental Laboratory, This letter refers to the Department of Energy's (DOE) investigation of the facts and circumstances concerning work planning and work control deficiencies with regard to the replacement of hot cell exhaust ventilation filters at Test Reactor Area Building 632, Idaho National Engineering and Environmental Laboratory (INEEL). The result of these deficiencies was that

377

Preliminary Notice of Violation, Isotopes Idaho, Inc. - EA-2000-04 |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Isotopes Idaho, Inc. - EA-2000-04 Isotopes Idaho, Inc. - EA-2000-04 Preliminary Notice of Violation, Isotopes Idaho, Inc. - EA-2000-04 May 19, 2000 Preliminary Notice of Violation issued to International Isotopes Idaho, Inc., related to Work Planning and Control Deficiencies associated with Replacement of Exhaust Ventilation Filters at the Test Reactor Area Hot Cell Facility at the Idaho National Engineering and Environmental Laboratory, This letter refers to the Department of Energy's (DOE) investigation of the facts and circumstances concerning work planning and work control deficiencies with regard to the replacement of hot cell exhaust ventilation filters at Test Reactor Area Building 632, Idaho National Engineering and Environmental Laboratory (INEEL). The result of these deficiencies was that

378

New capability for isotopic mass tracking in pyroprocess simulation  

Science Conference Proceedings (OSTI)

In support of the Integral Fast Reactor fuel recycle demonstration project at Argonne's Hot Fuel Examination Facility-South (HFEF/S) facility, a new computational code package called PYRO has been developed. The basic PYRO code (version 1-1) models the atomic mass flows and phase compositions in the electrorefiner (pyrochemical reprocessing vessel). It has been extended in version 1-2 to include tracking of {approximately}800 isotopic masses, their radioactive decay, and related phenomena. In a demonstration simulation, the processing of 24 batches of spent Experimental Breeder Reactor II(EBR-II) U-10% Zr driver fuel (burnup {approximately}8%) containing 20 kg of uranium per batch was modeled.

Liaw, J.R.; Ackerman, J.P.

1989-01-01T23:59:59.000Z

379

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

Science Conference Proceedings (OSTI)

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

Donna Post Guillen

2008-06-01T23:59:59.000Z

380

ISOTOPE SEPARATION AND ISOTOPE EXCHANGE. A Bibliography with Abstracts  

SciTech Connect

The unclassified literature covering 2498 reports from 1907 through 1957 has been searched for isotopic exchange and isotepic separation reactions involving U and the lighter elements of the periodic chart through atomic number 30. From 1953 to 1957, all elements were included Numerous references to isotope properties, isotopic ratios, and kinetic isotope effects were included. This is a complete revision of TID-3036 (Revised) issued June 4, 1954. An author index is included. (auth)

Begun, G.M.

1959-10-28T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS  

SciTech Connect

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Many research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.

Nichols, T.; Beals, D.; Sternat, M.

2011-07-18T23:59:59.000Z

382

REACTOR UNLOADING  

DOE Patents (OSTI)

This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

Leverett, M.C.

1958-02-18T23:59:59.000Z

383

NUCLEAR REACTOR  

DOE Patents (OSTI)

A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

Treshow, M.

1958-08-19T23:59:59.000Z

384

Brookhaven Graphite Research Reactor Workshop | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Services » Site & Facility Restoration » Deactivation & Services » Site & Facility Restoration » Deactivation & Decommissioning (D&D) » D&D Workshops » Brookhaven Graphite Research Reactor Workshop Brookhaven Graphite Research Reactor Workshop The Brookhaven Graphite Research Reactor (BGRR) was the first reactor built in the U.S. for peacetime atomic research following World War II. Construction began in 1947 and the reactor started operating in August 1950. In the next 18 years, an estimated 25,000 scientific experiments were carried out at the BGRR using neutrons produced in the facility's 700-ton graphite core, made up of more than 60,000 individual graphite blocks. The BGRR was placed on standby in 1968 and then permanently shut down as the next-generation reactor, the High Flux Beam Reactor (HFBR), was

385

Neutronic reactor  

DOE Patents (OSTI)

A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

Lewis, Warren R. (Richland, WA)

1978-05-30T23:59:59.000Z

386

NUCLEAR REACTORS  

DOE Patents (OSTI)

An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

1961-12-01T23:59:59.000Z

387

REACTOR CONTROL  

DOE Patents (OSTI)

This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

Ruano, W.J.

1957-12-10T23:59:59.000Z

388

Light Water Reactors Technology Development - Nuclear Reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

Light Water Reactors Light Water Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

389

1 Managed by UT-Battelle for the U.S. Department of Energy  

E-Print Network (OSTI)

Reactor (HFIR) will make ORNL the world's foremost center for neutron scattering. The Leadership Computing source. High Flux Isotope Reactor (HFIR): world's highest-flux reactor based neutron source. Leadership

390

Electroslag remelting with used fluxes  

Science Conference Proceedings (OSTI)

The Ukranian Scientific-Research Institute of Specialty Steel collaborated with plants engaged in the production of quality metals to introduce a low-waste electroslag remelting (ESR) technology employing used fluxes. It was established that the fluoride (type ANF-1) and fluoride-oxide (type ANF-6) fluxes which are widely used in ESR still have a high content of calcium fluoride and alumina and a low impurity content after 8-10 h of ESR. In the ESR of steels with used fluxes, the content of monitored components in the final slags changes negligibly, while the content of most impurities decreases. The used flux is also characterized by a low concentration of phosphorus and sulfur. It was found that flux can be used 3-5 times when it makes up 50% of the flux mixture in the charge. The savings realized from the use of spent flux in ESR amounts to 4-9 rubles/ton steel.

Yakovlev, N.F.; Sokha, Yu.S.; Oleinik, Yu.S.; Prokhorov, A.N.; Ol'shanskaya, T.V.

1988-05-01T23:59:59.000Z

391

DEEP WATER ISOTOPIC CURRENT ANALYZER  

DOE Patents (OSTI)

A deepwater isotopic current analyzer, which employs radioactive isotopes for measurement of ocean currents at various levels beneath the sea, is described. The apparatus, which can determine the direction and velocity of liquid currents, comprises a shaft having a plurality of radiation detectors extending equidistant radially therefrom, means for releasing radioactive isotopes from the shaft, and means for determining the time required for the isotope to reach a particular detector. (AEC)

Johnston, W.H.

1964-04-21T23:59:59.000Z

392

Reanalysis of the Reactor Neutrino Anomaly  

E-Print Network (OSTI)

We reanalyze the reactor neutrino anomaly, wherein it is suggested that only about 94% of the emitted antineutrino flux was detected in short baseline experiments. We find that the form of the corrections that lead to the anomaly are very uncertain for the 30% of the flux that is determined by forbidden beta-decay transitions. This uncertainty was estimated in four ways and is larger than the size of the anomaly, and is unlikely to be reduced without accurate direct measurements of the antineutrino flux. Neutrino physics conclusions based on the original anomaly need to be revisited, as do oscillation analyses that assumed that the antineutrino flux is known to better than ~5%.

A. C. Hayes; J. L. Friar; G. T. Garvey; Guy Jonkmans

2013-09-17T23:59:59.000Z

393

Biology and Soft Matter | ornl.gov  

NLE Websites -- All DOE Office Websites (Extended Search)

Spallation Neutron Source (SNS) and the reactor-based High Flux Isotope Reactor (HFIR), at Oak Ridge National Laboratory (ORNL). Researchers have access to new...

394

Y-12's first Open House-September 2-3, 1967, part 2  

NLE Websites -- All DOE Office Websites (Extended Search)

work is done for the Laboratory. For instance, Y-12 fabricated major components for the Molten Salt Reactor Experiment, the High Flux Isotope Reactor, the DCX fusion experiment...

395

Neutron Science Facilities Operating Status | ORNL Neutron Sciences  

NLE Websites -- All DOE Office Websites (Extended Search)

Neutron Science Facilities Operating Status High Flux Isotope Reactor The reactor is currently operating at 100% power for fuel cycle 449. Spallation Neutron Source SNS is shutdown...

396

Solar Thermal Reactor Materials Characterization  

DOE Green Energy (OSTI)

Current research into hydrogen production through high temperature metal oxide water splitting cycles has created a need for robust high temperature materials. Such cycles are further enhanced by the use of concentrated solar energy as a power source. However, samples subjected to concentrated solar radiation exhibited lifetimes much shorter than expected. Characterization of the power and flux distributions representative of the High Flux Solar Furnace(HFSF) at the National Renewable Energy Laboratory(NREL) were compared to ray trace modeling of the facility. In addition, samples of candidate reactor materials were thermally cycled at the HFSF and tensile failure testing was performed to quantify material degradation. Thermal cycling tests have been completed on super alloy Haynes 214 samples and results indicate that maximum temperature plays a significant role in reduction of strength. The number of cycles was too small to establish long term failure trends for this material due to the high ductility of the material.

Lichty, P. R.; Scott, A. M.; Perkins, C. M.; Bingham, C.; Weimer, A. W.

2008-03-01T23:59:59.000Z

397

Method for separating boron isotopes  

SciTech Connect

A method of separating boron isotopes .sup.10 B and .sup.11 B by laser-induced selective excitation and photodissociation of BCl.sub.3 molecules containing a particular boron isotope. The photodissociation products react with an appropriate chemical scavenger and the reaction products may readily be separated from undissociated BCl.sub.3, thus effecting the desired separation of the boron isotopes.

Rockwood, Stephen D. (Los Alamos, NM)

1978-01-01T23:59:59.000Z

398

Advanced Test Reactor National Scientific User Facility  

Science Conference Proceedings (OSTI)

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

Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

399

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

SciTech Connect

The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called User’s Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. User’s week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.

John Jackson; Todd Allen; Frances Marshall; Jim Cole

2013-03-01T23:59:59.000Z

400

Raman spectroscopic and mass spectrometric investigations of the hydrogen isotopes and isotopically labelled methane  

Science Conference Proceedings (OSTI)

Suitable analytical methods must be tested and developed for monitoring the individual process steps within the fuel cycle of a fusion reactor and for tritium accountability. The utility of laser-Raman spectroscopy accompanied by mass spectrometry with an Omegatron was investigated using the analysis of all hydrogen isotopes and isotopically labeled methanes as an example. The Omegatron is useful for analyzing all hydrogen isotopes mixed with the stable helium isotopes. The application of this mass spectrometer were demonstrated by analyzing mixtures of deuterated methanes. In addition, it was employed to study the radiochemical Witzbach exchange reaction between tritium and methanes. A laser-Raman spectrometer was designed for analysis of tritium-containing gases and was built from individual components. A tritium-compatible, metal-sealed Raman cuvette having windows with good optical properties and additional means for measuring the stray light was first used successfully in this work. The Raman spectra of the hydrogen isotopes were acquired in the pure rotation mode and in the rotation-vibration mode and were used for on. The deuterated methanes were measured by Raman spectroscopy, the wavenumbers determined were assigned to the corresponding vibrations, and the wavenumbers for the rotational fine-structure were summarized in tables. The fundamental Vibrations of the deuterated methanes produced Witzbach reactions were detected and assigned. The fundamental vibrations of the molecules were obtained with Raman spectroscopy for the first time in this work. The @-Raman spectrometer assembled is well suited for the analysis of tritium- containing gases and is practical in combination with mass spectrometry using an Omegatron, for studying gases used in fusion.

Jewett, J.R., Fluor Daniel Hanford

1997-02-24T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

Axi-symmetrical flow reactor for .sup.196 Hg photochemical enrichment  

DOE Patents (OSTI)

The present invention is directed to an improved photochemical reactor useful for the isotopic enrichment of a predetermined isotope of mercury, especially, .sup.196 Hg. Specifically, two axi-symmetrical flow reactors were constructed according to the teachings of the present invention. These reactors improve the mixing of the reactants during the photochemical enrichment process, affording higher yields of the desired .sup.196 Hg product. Measurements of the variation of yield (Y) and enrichment factor (E) along the flow axis of these reactors indicates very substantial improvement in process uniformity compared to previously used photochemical reactor systems. In one preferred embodiment of the present invention, the photoreactor system was built such that the reactor chamber was removable from the system without disturbing the location of either the photochemical lamp or the filter employed therewith.

Grossman, Mark W. (Belmont, MA)

1991-01-01T23:59:59.000Z

402

Literature review of United States utilities computer codes for calculating actinide isotope content in irradiated fuel  

SciTech Connect

This paper reviews the accuracy and precision of methods used by United States electric utilities to determine the actinide isotopic and element content of irradiated fuel. After an extensive literature search, three key code suites were selected for review. Two suites of computer codes, CASMO and ARMP, are used for reactor physics calculations; the ORIGEN code is used for spent fuel calculations. They are also the most widely used codes in the nuclear industry throughout the world. Although none of these codes calculate actinide isotopics as their primary variables intended for safeguards applications, accurate calculation of actinide isotopic content is necessary to fulfill their function.

Horak, W.C.; Lu, Ming-Shih

1991-12-01T23:59:59.000Z

403

Nuclear reactor  

DOE Patents (OSTI)

A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

Pennell, William E. (Greensburg, PA); Rowan, William J. (Monroeville, PA)

1977-01-01T23:59:59.000Z

404

ELECTRONUCLEAR REACTOR  

DOE Patents (OSTI)

An electronuclear reactor is described in which a very high-energy particle accelerator is employed with appropriate target structure to produce an artificially produced material in commercial quantities by nuclear transformations. The principal novelty resides in the combination of an accelerator with a target for converting the accelerator beam to copious quantities of low-energy neutrons for absorption in a lattice of fertile material and moderator. The fertile material of the lattice is converted by neutron absorption reactions to an artificially produced material, e.g., plutonium, where depleted uranium is utilized as the fertile material.

Lawrence, E.O.; McMillan, E.M.; Alvarez, L.W.

1960-04-19T23:59:59.000Z

405

Photocatalytic reactor  

DOE Patents (OSTI)

A photocatalytic reactor for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane.

Bischoff, Brian L. (Knoxville, TN); Fain, Douglas E. (Oak Ridge, TN); Stockdale, John A. D. (Knoxville, TN)

1999-01-01T23:59:59.000Z

406

Strengthening the nuclear-reactor fuel cycle against proliferation  

SciTech Connect

Argonne National Laboratory (ANL) conducts several research programs that serve to reduce the risks of fissile-material diversion from the nuclear-reactor fuel cycle. The objectives are to provide economical and efficient neutron or power generation with the minimum of inherent risks, and to further minimize risks by utilizing sophisticated techniques to detect attempts at material diversion. This paper will discuss the Reduced Enrichment Research and Test Reactor (RERTR) Program, the Isotope Correlation Technique (ICT), and Proliferation-Resistant Closed-Cycle Reactors. The first two are sponsored by the DOE Office of Arms Control and Nonproliferation.

Travelli, A.; Snelgrove, J.; Persiani, P. [Argonne National Lab., IL (United States). Arms Control and Nonproliferation Program

1992-12-31T23:59:59.000Z

407

Tritium Formation and Mitigation in High Temperature Reactors  

SciTech Connect

Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. In order to prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for the production of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750°C. Results of the diffusion model are presented for one steadystate value of tritium production in the reactor.

Piyush Sabharwall; Carl Stoots

2012-08-01T23:59:59.000Z

408

Advanced Online Flux Mapping of CANDU PHWR by Least-Squares Method  

Science Conference Proceedings (OSTI)

A least-squares method that solves both the core neutronics design equations and the in-core detector response equations on the least-squares principle is presented as a new advanced online flux-mapping method for CANada Deuterium Uranium (CANDU) pressurized heavy water reactors (PHWRs). The effectiveness of the new flux-mapping method is examined in terms of online flux-mapping calculations with numerically simulated true flux distribution and detector signals and those with the actual core-follow data for the Wolsong CANDU PHWRs in Korea. The effects of core neutronics models as well as the detector failures and uncertainties of measured detector signals on the effectiveness of the least-squares flux-mapping calculations are also examined.The following results are obtained. The least-squares method predicts the flux distribution in better agreement with the simulated true flux distribution than the standard core neutronics calculations by the finite difference method (FDM) computer code without using the detector signals. The adoption of the nonlinear nodal method based on the unified nodal method formulation instead of the FDM results in a significant improvement in prediction accuracy of the flux-mapping calculations. The detector signals estimated from the least-squares flux-mapping calculations are much closer to the measured detector signals than those from the flux synthesis method (FSM), the current online flux-mapping method for CANDU reactors. The effect of detector failures is relatively small so that the plant can tolerate up to 25% of detector failures without seriously affecting the plant operation. The detector signal uncertainties aggravate accuracy of the flux-mapping calculations, yet the effects of signal uncertainties of the order of 1% standard deviation can be tolerable without seriously degrading the prediction accuracy of the least-squares method. The least-squares method is disadvantageous because it requires longer CPU time than the existing FSM. Considering ever-increasing computer speed and the improved operational safety margin of CANDU reactors gained by accurate flux-mapping calculations, however, it is concluded that the least-squares method presents an effective alternative to the existing flux-mapping method for CANDU reactors.

Hong, In Seob [Seoul National University (Korea, Republic of); Kim, Chang Hyo [Seoul National University (Korea, Republic of); Suk, Ho Chun [Korea Atomic Energy Research Institute (Korea, Republic of)

2005-07-15T23:59:59.000Z

409

Selection of Isotopes and Elements for Fuel Cycle Analysis  

Science Conference Proceedings (OSTI)

Fuel cycle system analysis simulations examine how the selection among fuel cycle options for reactors, fuel, separation, and waste management impact uranium ore utilization, waste masses and volumes, radiotoxicity, heat to geologic repositories, isotope-dependent proliferation resistance measures, and so forth. Previously, such simulations have tended to track only a few actinide and fission product isotopes, those that have been identified as important to a few criteria from the standpoint of recycled material or waste, taken as a whole. After accounting for such isotopes, the residual mass is often characterized as “fission product other” or “actinide other”. However, detailed assessment of separation and waste management options now require identification of key isotopes and residual mass for Group 1A/2A elements (Rb, Cs, Sr, Ba), inert gases (Kr, Xe), halogens (Br, I), lanthanides, transition metals, transuranic (TRU), uranium, actinide decay products. The paper explains the rationale for a list of 81 isotopes and chemical elements to better support separation and waste management assessment in dynamic system analysis models such as Verifiable Fuel Cycle Simulation (VISION)

Steven J. Piet

2009-04-01T23:59:59.000Z

410

ARM - Measurement - CO2 flux  

NLE Websites -- All DOE Office Websites (Extended Search)

: CO2 flux The rate of flow for carbon dioxide, a heavy, colorless greenhouse gas. Categories Atmospheric Carbon, Surface Properties Instruments The above measurement is...

411

Physics of String Flux Compactifications  

E-Print Network (OSTI)

We provide a qualitative review of flux compactifications of string theory, focusing on broad physical implications and statistical methods of analysis.

Frederik Denef; Michael R. Douglas; Shamit Kachru

2007-01-06T23:59:59.000Z

412

Nuclear Reactor Accidents  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactor Accidents The accidents at the Three Mile Island (TMI) and Chernobyl nuclear reactors have triggered particularly intense concern about radiation hazards. The TMI accident,...

413

Principles of Reactor Physics  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Reactor Physics M A Smith Argonne National Laboratory Nuclear Engineering Division Phone: 630-252-9747, Email: masmith@anl.gov Abstract: Nuclear reactor physics deals with...

414

STATEMENT OF CONSIDERATIONS Advance Test Reactor Class Waiver  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Advance Test Reactor Class Waiver Advance Test Reactor Class Waiver W(C)-2008-004 The Advanced Test Reactor (A TR) is a pressurized water test reactor at the Idaho National Laboratory (INL) that operates at low pressure and temperature. The ATR was originally designed to study the effects of intense radiation on reactor material and fuels . It has a "Four Leaf Clover" design that allows a diverse array of testing locations. The unique design allows for different flux in various locations and specialized systems also allow for certain experiments to be run at their own temperature and pressure. The U.S. Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007. This designation will allow the ATR to

415

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

Daniels, F.

1962-12-18T23:59:59.000Z

416

Silicon Isotopic Fractionation of CAI-like Vacuum Evaporation Residues  

SciTech Connect

Calcium-, aluminum-rich inclusions (CAIs) are often enriched in the heavy isotopes of magnesium and silicon relative to bulk solar system materials. It is likely that these isotopic enrichments resulted from evaporative mass loss of magnesium and silicon from early solar system condensates while they were molten during one or more high-temperature reheating events. Quantitative interpretation of these enrichments requires laboratory determinations of the evaporation kinetics and associated isotopic fractionation effects for these elements. The experimental data for the kinetics of evaporation of magnesium and silicon and the evaporative isotopic fractionation of magnesium is reasonably complete for Type B CAI liquids (Richter et al., 2002, 2007a). However, the isotopic fractionation factor for silicon evaporating from such liquids has not been as extensively studied. Here we report new ion microprobe silicon isotopic measurements of residual glass from partial evaporation of Type B CAI liquids into vacuum. The silicon isotopic fractionation is reported as a kinetic fractionation factor, {alpha}{sub Si}, corresponding to the ratio of the silicon isotopic composition of the evaporation flux to that of the residual silicate liquid. For CAI-like melts, we find that {alpha}{sub Si} = 0.98985 {+-} 0.00044 (2{sigma}) for {sup 29}Si/{sup 28}Si with no resolvable variation with temperature over the temperature range of the experiments, 1600-1900 C. This value is different from what has been reported for evaporation of liquid Mg{sub 2}SiO{sub 4} (Davis et al., 1990) and of a melt with CI chondritic proportions of the major elements (Wang et al., 2001). There appears to be some compositional control on {alpha}{sub Si}, whereas no compositional effects have been reported for {alpha}{sub Mg}. We use the values of {alpha}Si and {alpha}Mg, to calculate the chemical compositions of the unevaporated precursors of a number of isotopically fractionated CAIs from CV chondrites whose chemical compositions and magnesium and silicon isotopic compositions have been previously measured.

Knight, K; Kita, N; Mendybaev, R; Richter, F; Davis, A; Valley, J

2009-06-18T23:59:59.000Z

417

ARM - Measurement - Latent heat flux  

NLE Websites -- All DOE Office Websites (Extended Search)

govMeasurementsLatent heat flux govMeasurementsLatent heat flux ARM Data Discovery Browse Data Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Latent heat flux The time rate of flow for the specific enthalpy difference between two phases of a substance at the same temperature, typically water. Categories Surface Properties Instruments The above measurement is considered scientifically relevant for the following instruments. Refer to the datastream (netcdf) file headers of each instrument for a list of all available measurements, including those recorded for diagnostic or quality assurance purposes. ARM Instruments CO2FLX : Carbon Dioxide Flux Measurement Systems ECOR : Eddy Correlation Flux Measurement System EBBR : Energy Balance Bowen Ratio Station

418

Solar proton fluxes since 1956  

SciTech Connect

The fluxes of protons emitted during solar flares since 1956 were evaluated. The depth-versus-activity profiles of /sup 56/Co in several lunar rocks are consistent with the solar-proton fluxes detected by experiments on several satellites. Only about 20% of the solar-proton-induced activities of /sup 22/Na and /sup 55/Fe in lunar rocks from early Apollo missions were produced by protons emitted from the sun during solar cycle 20 (1965--1975). The depth-versus-activity data for these radionuclides in several lunar rocks were used to determine the fluxes of protons during solar cycle 19 (1954--1964). The average proton fluxes for cycle 19 are about five times those for both the last million years and for cycle 20. These solar-proton flux variations correlate with changes in sunspot activity.

Reedy, R.C.

1977-04-13T23:59:59.000Z

419

ARM - Measurement - Sensible heat flux  

NLE Websites -- All DOE Office Websites (Extended Search)

govMeasurementsSensible heat flux govMeasurementsSensible heat flux ARM Data Discovery Browse Data Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Sensible heat flux The time rate of flow for the energy transferred from a warm or hot surface to whatever is touching it, typically air. Categories Surface Properties Instruments The above measurement is considered scientifically relevant for the following instruments. Refer to the datastream (netcdf) file headers of each instrument for a list of all available measurements, including those recorded for diagnostic or quality assurance purposes. ARM Instruments CO2FLX : Carbon Dioxide Flux Measurement Systems ECOR : Eddy Correlation Flux Measurement System EBBR : Energy Balance Bowen Ratio Station

420

Reactor and method of operation  

DOE Patents (OSTI)

A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

Wheeler, John A. (Princeton, NJ)

1976-08-10T23:59:59.000Z

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


421

Search for Neutrino Oscillations at the Palo Verde Nuclear Reactors  

E-Print Network (OSTI)

We report on the initial results from a measurement of the anti-neutrino flux and spectrum at a distance of about 800 m from the three reactors of the Palo Verde Nuclear Generating Station using a segmented gadolinium-loaded scintillation detector. We find that the anti-neutrino flux agrees with that predicted in the absence of oscillations to better than 5%, excluding at 90% CL $\\rm\\bar\

F. Boehm; J. Busenitz; B. Cook; G. Gratta; H. Henrikson; J. Kornis; D. Lawrence; K. B. Lee; K. McKinny; L. Miller; V. Novikov; A. Piepke; B. Ritchie; D. Tracy; P. Vogel; Y-F. Wang; J. Wolf

1999-12-22T23:59:59.000Z

422

Brookhaven Reactor Experiment Control Facility, a distributed function computer network  

SciTech Connect

A computer network for real-time data acquisition, monitoring and control of a series of experiments at the Brookhaven High Flux Beam Reactor has been developed and has been set into routine operation. This reactor experiment control facility presently services nine neutron spectrometers and one x-ray diffractometer. Several additional experiment connections are in progress. The architecture of the facility is based on a distributed function network concept. A statement of implementation and results is presented. (auth)

Dimmler, D.G.; Greenlaw, N.; Kelley, M.A.; Potter, D.W.; Rankowitz, S.; Stubblefield, F.W.

1975-11-01T23:59:59.000Z

423

Preserving physics knowledge at the fast flux test facility  

SciTech Connect

One of the goals of the Dept. of Energy's Office of Nuclear Energy, initiated under the Fuel Cycle Research and Development Program (FCRD) and continued under the Advanced Reactor Concepts Program (ARC) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMRs) that could support the development of an environmentally and economically sound nuclear fuel cycle. The Fast Flux Test Facility (FFTF) is the most recent LMR to operate in the United States, from 1982 to 1992, and was designed as a fully instrumented test reactor with on-line, real time test control and performance monitoring of components and tests installed in the reactor. The 10 years of operation of the FFTF provided a very useful framework for testing the advances in LMR safety technology based on passive safety features that may be of increased importance to new designs after the events at Fukushima. Knowledge preservation at the FFTF is focused on the areas of design, construction, and startup of the reactor, as well as on preserving information obtained from 10 years of successful operating history and extensive irradiation testing of fuels and materials. In order to ensure protection of information at risk, the program to date has sequestered reports, files, tapes, and drawings to allow for secure retrieval. A disciplined and orderly approach has been developed to respond to client's requests for documents and data in order to minimize the search effort and ensure that future requests for this information can be readily accommodated. (authors)

Wootan, D.; Omberg, R. [Pacific Northwest National Laboratory, 902 Battelle Boulevard, Richland, WA 99352 (United States); Makenas, B. J. [Ares Corporation, M/S A3-06, 825 Jadwin Avenue, Richland, WA 99352 (United States); Nielsen, D. L.; Nelson, J. V. [Indian Eyes, LLC, 2815 Saint Andrews Loop, Pasco, WA 99301 (United States); Polzin, D. L. [CH2MHill Plateau Remediation Company, M/S S2-42, P.O. Box 1600, Richland, WA 99352 (United States)

2012-07-01T23:59:59.000Z

424

u---lm---n piltering with snequ---lity gonstr---ints for ''ur~of---n ingine re---lth istim---tion  

E-Print Network (OSTI)

A Proposal: Reliability Centered Maintenance (RCM) for the High Flux Isotope Reactor (HFIR) Eugene Isotope Reactor (HFIR). The PCS has shown decreasing reliability and performance over the last several-centered maintenance at the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory to improve system

Simon, Dan

425

Spectral functions and zeta functions in hyperbolic spaces Roberto Camporesi and Atsushi Higuchi  

E-Print Network (OSTI)

A Proposal: Reliability Centered Maintenance (RCM) for the High Flux Isotope Reactor (HFIR) Eugene Isotope Reactor (HFIR). The PCS has shown decreasing reliability and performance over the last several-centered maintenance at the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory to improve system

Ceragioli, Francesca

426

Gamma displacement cross-sections in various materials  

E-Print Network (OSTI)

A Proposal: Reliability Centered Maintenance (RCM) for the High Flux Isotope Reactor (HFIR) Eugene Isotope Reactor (HFIR). The PCS has shown decreasing reliability and performance over the last several-centered maintenance at the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory to improve system

Motta, Arthur T.

427

Isotopically labeled compositions and method  

DOE Patents (OSTI)

Compounds having stable isotopes .sup.13C and/or .sup.2H were synthesized from precursor compositions having solid phase supports or affinity tags.

Schmidt, Jurgen G. (Los Alamos, NM); Kimball, David B. (Los Alamos, NM); Alvarez, Marc A. (Santa Fe, NM); Williams, Robert F. (Los Alamos, NM); Martinez, Rudolfo A. (Santa Fe, NM)

2011-07-12T23:59:59.000Z

428

Prospects for Tokamak Fusion Reactors  

SciTech Connect

This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.

Sheffield, J.; Galambos, J.

1995-04-01T23:59:59.000Z

429

Reactor safety method  

DOE Patents (OSTI)

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

Vachon, Lawrence J. (Clairton, PA)

1980-03-11T23:59:59.000Z

430

NEUTRONIC REACTOR MANIPULATING DEVICE  

DOE Patents (OSTI)

A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

Ohlinger, L.A.

1962-08-01T23:59:59.000Z

431

The boardman regional flux experiment  

SciTech Connect

A field campaign was carried out near Boardman, Oregon, to study the effects of subgrid-scale variability of sensible- and latent-heat fluxes on surface boundary-layer properties. The experiment involved three U.S. Department of Energy laboratories, one National Oceanic and Atmospheric Administration laboratory, and several universities. The experiment was conducted in a region of severe contrasts in adjacent surface types that accentuated the response of the atmosphere to variable surface forcing. Large values of sensible-heat flux and low values of latent-heat flux characterized a sagebrush steppe area; significantly smaller sensible-heat fluxes and much larger latent-heat fluxes were associated with extensive tracts of irrigated farmland to the north, east, and west of the steppe. Data were obtained from an array of surface flux stations, remote-sensing devices, an instrumented aircraft, and soil and vegetation measurements. The data will be used to address the problem of extrapolating from a limited number of local measurements to area-averaged values of fluxes suitable for use in global climate models. 16 refs., 13 figs.

Doran, J.C.; Hubbe, J.M.; Kirkham, R.R.; Shaw, W.J.; Whiteman, C.D. (Pacific Northwest Lab., Richland, WA (United States)); Barnes, F.J.; Cooper, D.; Porch, W. (Los Alamos National Lab., NM (United States)); Coulter, R.L.; Cook, D.R.

1992-11-01T23:59:59.000Z

432

Improvement of Pt/C/PTFE catalyst type used for hydrogen isotope separation  

Science Conference Proceedings (OSTI)

The CANDU reactor from the Nuclear Power plant Cernavoda (Romania)) is the most powerful tritium source from Europe. This reactor is moderated and cooled by heavy water that becomes continuously contaminated with tritium. Because of this reason, the National R and amp;D Inst. for Cryogenic and Isotopic Technologies developed a detritiation technology based on catalytic isotopic exchange and cryogenic distillation. The main effort of our Inst. was focused on finding more efficient catalysts with a longer operational life. Some of the tritium removal processes involved in Fusion Science and Technology use this type of catalyst 1. Several Pt/C/PTFE hydrophobic catalysts that could be used in isotopic exchange process 2,3,4 were produced. The present paper presents a comparative study between the physical and morphological properties of different catalysts manufactured by impregnation at our institute. The comparison consists of a survey of specific surface, pores volume and pores distribution. (authors)

Vasut, F.; Preda, A.; Zamfirache, M.; Bornea, A. M.; Stefanescu, I.; Pearsica, C. [INC-DTCI, ICSI Rm. Valcea, Uzinei Street 4, Rm. Valcea (Romania)

2008-07-15T23:59:59.000Z

433

THE ADVANCED TEST REACTOR-ATR FINAL CONCEPTUAL DESIGN  

SciTech Connect

The results of a study are presented which provided additional experimental-loop irradiation space for the AECDRD testing program. It was a premise that the experiments allocated to this reactor were those which could not be accommodated in the MTR, ETR, or in existing commercial test reactors. To accomplish the design objectives called for a reactor producing perturbed neutron fluxes exceeding 1O/sup 15/ thermal n/cm/sup 2/-sec and 1.5 x 1O/sup 15/ epithermal n/cm/sup 2/-sec. To accommodate the experimental samples, the reactor fuel core is four feet long in the direction of experimental loops. This is twice the length of the MTR core and a third longer than the ETR core. The vertical arrangement of reactor and experiments permits the use of loops penetrating the top cap of the reactor vessel running straight and vertically through the reactor core. The design offers a high degree of accessibility of the exterior portions of the experiments and offers very convenient handling and discharge of experiments. Since the loops are to be integrated into the reactor design and the in-pile portions installed before reactor start-up, it is felt that many of the problems encountered in MTR and ETR experience will cease to exist. Installation of the loops prior to startup will have an added advantage in that the flux variations experienced in experiments in ETR every time a new loop is installed will be absent. The Advanced Test Reactor has a core configuration that provides essentially nine flux-trap regions in a geometry that is almost optimum for cylindrical experiments. The geometry is similar to that of a fourleaf clover with one flux trap in each leaf, one at the intersection of the leaves, and one between each pair of leaves. The nominal power level is 250 Mw. The study was carried out in enough detail to permit the establishment of the design parameters and to develop the power requirement which, conservatively rated, will definitely reach the flux specifications. A critical mockup of an arrangement similar to ATR was loaded into the Engineering Test Reactor Critical Facility. (auth)

deBoisblanc, D.R. et al

1960-11-01T23:59:59.000Z

434

ISOTOPE FRACTIONATION PROCESS  

DOE Patents (OSTI)

A new method is described for isotopic enrichment of uranium. It has been found that when an aqueous acidic solution of ionic tetravalent uraniunn is contacted with chelate complexed tetravalent uranium, the U/sup 238/ preferentially concentrates in the complexed phase while U/sup 235/ concentrates in the ionic phase. The effect is enhanced when the chelate compound is water insoluble and is dissolved in a water-immiscible organic solvent. Cupferron is one of a number of sultable complexing agents, and chloroform is a suitable organic solvent.

Clewett, G.H.; Lee, DeW.A.

1958-05-20T23:59:59.000Z

435

An Engineering Test Reactor  

SciTech Connect

A relatively inexpensive reactor for the specific purpose of testing a sub-critical portion of another reactor under conditions that would exist during actual operation is discussed. It is concluded that an engineering tool for reactor development work that bridges the present gap between exponential and criticality experiments and the actual full scale operating reactor is feasible. An example of such a test reactor which would not entail development effort to ut into operation is depicted.

Fahrner, T.; Stoker, R.L.; Thomson, A.S.

1951-03-16T23:59:59.000Z

436

Fast Flux Test Facility (FFTF) standby plan  

Science Conference Proceedings (OSTI)

The FFTF Standby Plan, Revision 0, provides changes to the major elements and project baselines to maintain the FFTF plant in a standby condition and to continue washing sodium from irradiated reactor fuel. The Plan is consistent with the Memorandum of Decision approved by the Secretary of Energy on January 17, 1997, which directed that FFTF be maintained in a standby condition to permit the Department to make a decision on whether the facility should play a future role in the Department of Energy`s dual track tritium production strategy. This decision would be made in parallel with the intended December 1998 decision on the selection of the primary, long- term source of tritium. This also allows the Department to review the economic and technical feasibility of using the FFTF to produce isotopes for the medical community. Formal direction has been received from DOE-RL and Fluor 2020 Daniel Hanford to implement the FFTF standby decision. The objective of the Plan is maintain the condition of the FFTF systems, equipment and personnel to preserve the option for plant restart within three and one-half years of a decision to restart, while continuing deactivation work which is consistent with the standby mode.

Hulvey, R.K.

1997-03-06T23:59:59.000Z

437

Reactor Pressure Vessel Task of Light Water Reactor Sustainability...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure...

438

ARM - Measurement - Soil heat flux  

NLE Websites -- All DOE Office Websites (Extended Search)

heat flux heat flux ARM Data Discovery Browse Data Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Soil heat flux A quantity measured according to the formula B = {lambda}(dT/dz), where {lambda} is the conductivity of the soil that the heat is moving through. Categories Surface Properties Instruments The above measurement is considered scientifically relevant for the following instruments. Refer to the datastream (netcdf) file headers of each instrument for a list of all available measurements, including those recorded for diagnostic or quality assurance purposes. ARM Instruments CO2FLX : Carbon Dioxide Flux Measurement Systems EBBR : Energy Balance Bowen Ratio Station SEBS : Surface Energy Balance System External Instruments

439

Estimation of Surface Heat Flux  

Science Conference Proceedings (OSTI)

The authors reconsider the problem of estimating the sensible heat transfer at the earth's surface from direct measurements of turbulent fluxes in the atmospheric boundary layer. For simplicity, only horizontally homogeneous conditions are ...

Jielun Sun; Steven K. Esbensen; L. Mahrt

1995-09-01T23:59:59.000Z

440

ARM - Measurement - Soil moisture flux  

NLE Websites -- All DOE Office Websites (Extended Search)

moisture flux moisture flux ARM Data Discovery Browse Data Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Soil moisture flux A quantity measured according to the formula B = {lambda}(dq/dz), where {lambda} is the conductivity of the soil that the moisture is moving through. Categories Surface Properties Instruments The above measurement is considered scientifically relevant for the following instruments. Refer to the datastream (netcdf) file headers of each instrument for a list of all available measurements, including those recorded for diagnostic or quality assurance purposes. ARM Instruments CO2FLX : Carbon Dioxide Flux Measurement Systems External Instruments ECMWFDIAG : European Centre for Medium Range Weather Forecasts

Note: This page contains sample records for the topic "flux isotope reactor" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
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441

Flux Measurement with Conditional Sampling  

Science Conference Proceedings (OSTI)

A method is proposed to measure scalar fluxes using conditional sampling. Only the mean concentrations of updraft and downdraft samples, the standard deviation of the vertical velocity, and a coefficient of proportionality, b, need to be known. ...

Joost A. Businger; Steven P. Oncley

1990-04-01T23:59:59.000Z

442

The Solar Wind Energy Flux  

E-Print Network (OSTI)

The solar-wind energy flux measured near the ecliptic is known to be independent of the solar-wind speed. Using plasma data from Helios, Ulysses, and Wind covering a large range of latitudes and time, we show that the solar-wind energy flux is independent of the solar-wind speed and latitude within 10%, and that this quantity varies weakly over the solar cycle. In other words the energy flux appears as a global solar constant. We also show that the very high speed solar-wind (VSW > 700 km/s) has the same mean energy flux as the slower wind (VSW solar-wind speed and density, which formalizes the anti-correlation between these quantities.

Chat, G Le; Meyer-Vernet, N

2012-01-01T23:59:59.000Z

443

Grid-Averaged Surface Fluxes  

Science Conference Proceedings (OSTI)

This study examines the inadequacies of formulations for surface fluxes for use in numerical models of atmospheric flow. The difficulty is that numerical models imply spatial averaging over each grid area. Existing formulations am based on the ...

L. Mahrt

1987-08-01T23:59:59.000Z

444

Method of separating boron isotopes  

SciTech Connect

A method of boron isotope enrichment involving the isotope preferential photolysis of (2-chloroethenyl)dichloroborane as the feed material. The photolysis can readily be achieved with CO.sub.2 laser radiation and using fluences significantly below those required to dissociate BCl.sub.3.

Jensen, Reed J. (Los Alamos, NM); Thorne, James M. (Provo, UT); Cluff, Coran L. (Provo, UT); Hayes, John K. (Salt Lake City, UT)

1984-01-01T23:59:59.000Z

445

Prospects For Precision Measurements with Reactor Antineutrinos at Daya Bay  

E-Print Network (OSTI)

In 2012 the Daya Bay experiment made an unambiguous observation of reactor antineutrino disappearance over kilometer-long baselines and determined that the neutrino mixing angle $\\theta_{13}$ is non-zero. The measurements of Daya Bay have provided the most precise determination of $\\theta_{13}$ to date. This whitepaper outlines the prospects for precision studies of reactor antineutrinos at Daya Bay in the coming years. This includes precision measurements of sin$^2 2\\theta_{13}$ and $\\Delta m^2_{ee}$ to $reactor flux and spectrum, and non-standard physics searches.

The Daya Bay Collaboration

2013-09-30T23:59:59.000Z

446

LIGHT WATER MODERATED NEUTRONIC REACTOR  

DOE Patents (OSTI)

A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

Christy, R.F.; Weinberg, A.M.

1957-09-17T23:59:59.000Z

447

Definition: Gas Flux Sampling | Open Energy Information  

Open Energy Info (EERE)

Gas Flux Sampling Jump to: navigation, search Dictionary.png Gas Flux Sampling Gas flux sampling measures the flow of volatile gas emissions from a specific location and compares...

448

Flux Sampling Errors for Aircraft and Towers  

Science Conference Proceedings (OSTI)

Various errors and influences leading to differences between tower- and aircraft-measured fluxes are surveyed. This survey is motivated by reports in the literature that aircraft fluxes are sometimes smaller than tower-measured fluxes. Both tower ...

L. Mahrt

1998-04-01T23:59:59.000Z

449

Isotopic Analysis- Fluid | Open Energy Information  

Open Energy Info (EERE)

Isotopic Analysis- Fluid Isotopic Analysis- Fluid Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Technique: Isotopic Analysis- Fluid Details Activities (61) Areas (32) Regions (6) NEPA(0) Exploration Technique Information Exploration Group: Lab Analysis Techniques Exploration Sub Group: Fluid Lab Analysis Parent Exploration Technique: Fluid Lab Analysis Information Provided by Technique Lithology: Water rock interaction Stratigraphic/Structural: Hydrological: Origin of hydrothermal fluids; Mixing of hydrothermal fluids Thermal: Isotopic ratios can be used to characterize and locate subsurface thermal anomalies. Dictionary.png Isotopic Analysis- Fluid: Isotopes are atoms of the same element that have different numbers of neutrons. An isotopic analysis looks at a particular isotopic element(s) in

450

It's Elemental - Isotopes of the Element Neptunium  

NLE Websites -- All DOE Office Websites (Extended Search)

Uranium Previous Element (Uranium) The Periodic Table of Elements Next Element (Plutonium) Plutonium Isotopes of the Element Neptunium Click for Main Data Most of the isotope...

451

It's Elemental - Isotopes of the Element Sulfur  

NLE Websites -- All DOE Office Websites (Extended Search)

Phosphorus Previous Element (Phosphorus) The Periodic Table of Elements Next Element (Chlorine) Chlorine Isotopes of the Element Sulfur Click for Main Data Most of the isotope...

452

It's Elemental - Isotopes of the Element Argon  

NLE Websites -- All DOE Office Websites (Extended Search)

Chlorine Previous Element (Chlorine) The Periodic Table of Elements Next Element (Potassium) Potassium Isotopes of the Element Argon Click for Main Data Most of the isotope data...

453

The marine biogeochemistry of zinc isotopes  

E-Print Network (OSTI)

Zinc (Zn) stable isotopes can record information about important oceanographic processes. This thesis presents data on Zn isotopes in anthropogenic materials, hydrothermal fluids and minerals, cultured marine phytoplankton, ...

John, Seth G

2007-01-01T23:59:59.000Z

454