National Library of Energy BETA

Sample records for helium reactor gt-mhr

  1. Evaluation of the Gas Turbine Modular Helium Reactor

    SciTech Connect (OSTI)

    Not Available

    1994-02-01

    Recent advances in gas-turbine and heat exchanger technology have enhanced the potential for a Modular Helium Reactor (MHR) incorporating a direct gas turbine (Brayton) cycle for power conversion. The resulting Gas Turbine Modular Helium Reactor (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. While the passive safety features of the steam cycle MHR (SC-MHR) are retained, generation efficiencies are projected to be in the range of 48% and steam power conversion systems, with their attendant complexities, are eliminated. Power costs are projected to be reduced by about 20%, relative to the SC-MHR or coal. This report documents the second, and final, phase of a two-part evaluation that concluded with a unanimous recommendation that the direct cycle (DC) variant of the GT-MHR be established as the commercial objective of the US Gas-Cooled Reactor Program. This recommendation has been endorsed by industrial and utility participants and accepted by the US Department of Energy (DOE). The Phase II effort, documented herein, concluded that the DC GT-MHR offers substantial technical and economic advantages over both the IDC and SC systems. Both the DC and IDC were found to offer safety advantages, relative to the SC, due to elimination of the potential for water ingress during power operations. This is the dominant consequence event for the SC. The IDC was judged to require somewhat less development than the direct cycle, while the SC, which has the greatest technology base, incurs the least development cost and risk. While the technical and licensing requirements for the DC were more demanding, they were judged to be incremental and feasible. Moreover, the DC offers significant performance and cost improvements over the other two concepts. Overall, the latter were found to justify the additional development needs.

  2. Computational Assessment of the GT-MHR Graphite Core Support Structural Integrity in Air-Ingress Accident Condition

    SciTech Connect (OSTI)

    Jong B. Lim; Eung S. Kim; Chang H. Oh; Richard R. Schultz; David A. Petti

    2008-10-01

    The objective of this project was to perform stress analysis for graphite support structures of the General Atomics’ 600 MWth GT-MHR prismatic core design using ABAQUS ® (ver. 6.75) to assess their structural integrity in air-ingress accident conditions where the structure weakens over time due to oxidation damages. The graphite support structures of prismatic type GT-MHR was analyzed based on the change of temperature, burn-off and corrosion depth during the accident period predicted by GAMMA, a multi-dimensional gas multi-component mixture analysis code developed in the Republic of Korea (ROK)/United States (US) International –Nuclear Engineering Research Initiative (I-NERI) project. Both the loading and thermal stresses were analyzed, but the thermal stress was not significant, leaving the loading stress to be the major factor. The mechanical strengths are exceeded between 11 to 11.5 days after loss-of-coolant-accident (LOCA), corresponding to 5.5 to 6 days after the start of natural convection.

  3. Deep-Burn Modular Helium Reactor Fuel Development Plan

    SciTech Connect (OSTI)

    McEachern, D

    2002-12-02

    This document contains the workscope, schedule and cost for the technology development tasks needed to satisfy the fuel and fission product transport Design Data Needs (DDNs) for the Gas Turbine-Modular Helium Reactor (GT-MHR), operating in its role of transmuting transuranic (TRU) nuclides in spent fuel discharged from commercial light-water reactors (LWRs). In its application for transmutation, the GT-MHR is referred to as the Deep-Burn MHR (DB-MHR). This Fuel Development Plan (FDP) describes part of the overall program being undertaken by the U.S. Department of Energy (DOE), utilities, and industry to evaluate the use of the GT-MHR to transmute transuranic nuclides from spent nuclear fuel. The Fuel Development Plan (FDP) includes the work on fuel necessary to support the design and licensing of the DB-MHR. The FDP is organized into ten sections. Section 1 provides a summary of the most important features of the plan, including cost and schedule information. Section 2 describes the DB-MHR concept, the features of its fuel and the plan to develop coated particle fuel for transmutation. Section 3 describes the knowledge base for fabrication of coated particles, the experience with irradiation performance of coated particle fuels, the database for fission product transport in HTGR cores, and describes test data and calculations for the performance of coated particle fuel while in a repository. Section 4 presents the fuel performance requirements in terms of as-manufactured quality and performance of the fuel coatings under irradiation and accident conditions. These requirements are provisional because the design of the DB-MHR is in an early stage. However, the requirements are presented in this preliminary form to guide the initial work on the fuel development. Section 4 also presents limits on the irradiation conditions to which the coated particle fuel can be subjected for the core design. These limits are based on past irradiation experience. Section 5 describes

  4. LANL researchers simulate helium bubble behavior in fusion reactors

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Researchers simulate helium bubble behavior LANL researchers simulate helium bubble behavior in fusion reactors A team performed simulations to understand more fully how tungsten behaves in such harsh conditions, particularly in the presence of implanted helium that forms bubbles in the material. August 4, 2015 Simulation snapshots of the helium bubble just before bursting. Colors indicate tungsten atoms (red) and helium atoms (blue). Simulation snapshots of the helium bubble just before

  5. Very High Temperature Reactor (VHTR) Survey of Materials Research and Development Needs to Support Early Deployment

    SciTech Connect (OSTI)

    Eric Shaber; G. Baccaglini; S. Ball; T. Burchell; B. Corwin; T. Fewell; M. Labar; P. MacDonald; P. Rittenhouse; Russ Vollam; F. Southworth

    2003-01-01

    The VHTR reference concept is a helium-cooled, graphite moderated, thermal neutron spectrum reactor with an outlet temperature of 1000 C or higher. It is expected that the VHTR will be purchased in the future as either an electricity producing plant with a direct cycle gas turbine or a hydrogen producing (or other process heat application) plant. The process heat version of the VHTR will require that an intermediate heat exchanger (IHX) and primary gas circulator be located in an adjoining power conversion vessel. A third VHTR mission - actinide burning - can be accomplished with either the hydrogen-production or gas turbine designs. The first ''demonstration'' VHTR will produce both electricity and hydrogen using the IHX to transfer the heat to either a hydrogen production plant or the gas turbine. The plant size, reactor thermal power, and core configuration will be designed to assure passive decay heat removal without fuel damage during accidents. The fuel cycle will be a once-through very high burnup low-enriched uranium fuel cycle. The purpose of this report is to identify the materials research and development needs for the VHTR. To do this, we focused on the plant design described in Section 2, which is similar to the GT-MHR plant design (850 C core outlet temperature). For system or component designs that present significant material challenges (or far greater expense) there may be some viable design alternatives or options that can reduce development needs or allow use of available (cheaper) materials. Nevertheless, we were not able to assess those alternatives in the time allotted for this report and, to move forward with this material research and development assessment, the authors of this report felt that it was necessary to use a GT-MHR type design as the baseline design.

  6. Fuel Development For Gas-Cooled Fast Reactors

    SciTech Connect (OSTI)

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic honeycomb structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  7. Liquid uranium alloy-helium fission reactor

    DOE Patents [OSTI]

    Minkov, V.

    1984-06-13

    This invention describes a nuclear fission reactor which has a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200 to 1800/sup 0/C range, and even higher to 2500/sup 0/C.

  8. Liquid uranium alloy-helium fission reactor

    DOE Patents [OSTI]

    Minkov, Vladimir (Skokie, IL)

    1986-01-01

    This invention teaches a nuclear fission reactor having a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200.degree.-1800.degree. C. range, and even higher to 2500.degree. C., limited only by the thermal effectiveness of the structural materials, increasing the efficiency of power generation from the normal 30-35% with 300.degree.-500.degree. C. upper limit temperature to 50-65%. Irradiation of the circulating liquid fuel, as contrasted to only localized irradiation of a solid fuel, provides improved fuel utilization.

  9. Helium bubble distributions in reactor tank repair specimens. Part 1

    SciTech Connect (OSTI)

    Tosten, M.H.; Kestin, P.A.

    1992-03-01

    This report discusses the Reactor Tank Repair (RTR) program was initiated to develop an in-tank repair process capable of repairing stress corrosion cracks within the SRS reactor tank walls, in the event that such a repair is needed. Previous attempts to repair C-reactor tank with a gas tungsten arc (GTA) welding process were unsuccessful due to significant cracking that occurred in the heat-affected-zones adjacent to the repair welds. It was determined that this additional cracking was a result of helium embrittlement caused by the combined effects of helium (existing within the tank walls), the high heat input associated with the GTA process, and weld shrinkage stresses. Based on the results of earlier studies it was suggested that the effects of helium embrittlement could be minimized by using a low heat input GMA process. Metallographic analysis played an important role throughout the investigation of alternative welding methods for the repair of helium-containing materials.

  10. Helium bubble distributions in reactor tank repair specimens

    SciTech Connect (OSTI)

    Tosten, M.H.; Kestin, P.A.

    1992-03-01

    This report discusses the Reactor Tank Repair (RTR) program was initiated to develop an in-tank repair process capable of repairing stress corrosion cracks within the SRS reactor tank walls, in the event that such a repair is needed. Previous attempts to repair C-reactor tank with a gas tungsten arc (GTA) welding process were unsuccessful due to significant cracking that occurred in the heat-affected-zones adjacent to the repair welds. It was determined that this additional cracking was a result of helium embrittlement caused by the combined effects of helium (existing within the tank walls), the high heat input associated with the GTA process, and weld shrinkage stresses. Based on the results of earlier studies it was suggested that the effects of helium embrittlement could be minimized by using a low heat input GMA process. Metallographic analysis played an important role throughout the investigation of alternative welding methods for the repair of helium-containing materials.

  11. Multiple reheat helium Brayton cycles for sodium fast reactors

    SciTech Connect (OSTI)

    Haihua Zhao; Per F. Peterson

    2008-07-01

    Sodium fast reactors (SFR) traditionally adopt the steam Rankine cycle for power conversion. The resulting potential for water-sodium reaction remains a continuing concern which at least partly delays the SFR technology commercialization and is a contributor to higher capital cost. Supercritical CO2 provides an alternative, but is also capable of sustaining energetic chemical reactions with sodium. Recent development on advanced inert-gas Brayton cycles could potentially solve this compatibility issue, increase thermal efficiency, and bring down the capital cost close to light water reactors. In this paper, helium Brayton cycles with multiple reheat and intercooling states are presented for SFRs with reactor outlet temperatures in the range of 510°C to 650°C. The resulting thermal efficiencies range from 39% and 47%, which is comparable with supercritical recompression CO2 cycles (SCO2 cycle). A systematic comparison between multiple reheat helium Brayton cycle and the SCO2 cycle is given, considering compatibility issues, plant site cooling temperature effect on plant efficiency, full plant cost optimization, and other important factors. The study indicates that the multiple reheat helium cycle is the preferred choice over SCO2 cycle for sodium fast reactors.

  12. Investigation of Countercurrent Helium-Air Flows in Air-ingress Accidents for VHTRs

    SciTech Connect (OSTI)

    Sun, Xiaodong; Christensen, Richard; Oh, Chang

    2013-10-03

    The primary objective of this research is to develop an extensive experimental database for the air- ingress phenomenon for the validation of computational fluid dynamics (CFD) analyses. This research is intended to be a separate-effects experimental study. However, the project team will perform a careful scaling analysis prior to designing a scaled-down test facility in order to closely tie this research with the real application. As a reference design in this study, the team will use the 600 MWth gas turbine modular helium reactor (GT-MHR) developed by General Atomic. In the test matrix of the experiments, researchers will vary the temperature and pressure of the helium— along with break size, location, shape, and orientation—to simulate deferent scenarios and to identify potential mitigation strategies. Under support of the Department of Energy, a high-temperature helium test facility has been designed and is currently being constructed at Ohio State University, primarily for high- temperature compact heat exchanger testing for the VHTR program. Once the facility is in operation (expected April 2009), this study will utilize high-temperature helium up to 900°C and 3 MPa for loss-of-coolant accident (LOCA) depressurization and air-ingress experiments. The project team will first conduct a scaling study and then design an air-ingress test facility. The major parameter to be measured in the experiments is oxygen (or nitrogen) concentration history at various locations following a LOCA scenario. The team will use two measurement techniques: 1) oxygen (or similar type) sensors employed in the flow field, which will introduce some undesirable intrusiveness, disturbing the flow, and 2) a planar laser-induced fluorescence (PLIF) imaging technique, which has no physical intrusiveness to the flow but requires a transparent window or test section that the laser beam can penetrate. The team will construct two test facilities, one for high-temperature helium tests with

  13. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    SciTech Connect (OSTI)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  14. Remote reactor repair: GTA (gas tungsten Arc) weld cracking caused by entrapped helium

    SciTech Connect (OSTI)

    Kanne, W.R. Jr.

    1988-01-01

    A repair patch was welded to the wall of a nuclear reactor tank using remotely controlled thirty-foot long robot arms. Further repair was halted when gas tungsten arc (GTA) welds joining type 304L stainless steel patches to the 304 stainless steel wall developed toe cracks in the heat-affected zone (HAZ). The role of helium in cracking was investigated using material with entrapped helium from tritium decay. As a result of this investigation, and of an extensive array of diagnostic tests performed on reactor tank wall material, helium embrittlement was shown to be the cause of the toe cracks.

  15. Helium-cooled solid breeder blanket design for a tokamak fusion reactor

    SciTech Connect (OSTI)

    Huggenberger, M.; Schultz, K.R.

    1983-11-01

    A preliminary design for a helium-cooled solid breeder blanket for a tokamak fusion reactor has been developed, and its performance looks quite good. The design is capable of bearing a 4 MW/m/sup 2/ neutron wall load, and the ideal pumping power required for the whole primary helium loop including the steam generators is only 2.5% of the total thermal power. The maximum blanket thickness including the helium duct work is only 860 mm, the minimum thickness is only 730 mm. The design work was focused on the thermalhydraulic aspects, which represent the key problems associated with using helium as a coolant. The present work demonstrates that the potential disadvantages helium has, due to its limited heat transfer capabilities, can be avoided or minimized by an appropriate thermal-hydraulic design. As a result, helium with its many advantages remains a promising fusion blanket coolant.

  16. Material Control and Accounting Design Considerations for High-Temperature Gas Reactors

    SciTech Connect (OSTI)

    Trond Bjornard; John Hockert

    2011-08-01

    The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC&A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC&A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC&A approaches for the two major HTGR reactor types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR [Pty] and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC&A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR&D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present work summarized in

  17. Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

    SciTech Connect (OSTI)

    Middleton, Bobby; Pasch, James Jay; Kruizenga, Alan Michael; Walker, Matthew

    2016-01-01

    This report outlines the thermodynamics of a supercritical carbon dioxide (sCO2) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO2 code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related to both Helium and to sCO2 , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO2> Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO2 Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation

  18. Deep Burn Develpment of Transuranic Fuel for High-Temperature Helium-Cooled Reactors - July 2010

    SciTech Connect (OSTI)

    Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D; Bell, Gary L

    2010-08-01

    The DB Program Quarterly Progress Report for April - June 2010, ORNL/TM/2010/140, was distributed to program participants on August 4. This report discusses the following: (1) TRU (transuranic elements) HTR (high temperature helium-cooled reactor) Fuel Modeling - (a) Thermochemical Modeling, (b) 5.3 Radiation Damage and Properties; (2) TRU HTR Fuel Qualification - (a) TRU Kernel Development, (b) Coating Development, (c) ZrC Properties and Handbook; and (3) HTR Fuel Recycle - (a) Recycle Processes, (b) Graphite Recycle, (c) Pyrochemical Reprocessing - METROX (metal recovery from oxide fuel) Process Development.

  19. Optimal Coupling of a Nuclear Reactor and a Thermal Desalination Plant

    SciTech Connect (OSTI)

    Caruso, G.; Naviglio, A.; Nisan, S.; Bielak, B.; Cinotti, L.; Humphries, J.R.; Martins, N.; Volpi, L.

    2002-07-01

    The present study, performed in the framework of the EURODESAL Project (5. EU FWP), deals with the analysis of the 'optimum' coupling of a PWR and of a HTGR plant with a thermal desalination plant, based on the Multiple Effects process. The reference reactors are the AP600 and the PWR900 as Pressurized reactors and the GT-MHR as Gas reactor. The calculations performed show that there are several technical solutions allowing to couple PWRs and GRs to a ME desalination plant. The optimization criteria concern the technical feasibility of the coupling, producing the maximum quantity of fresh water at the lower cost, without unacceptable reduction of the electrical power produced and without undue health hazard for population. (authors)

  20. Improvements of fuel failure detection in boiling water reactors using helium measurements

    SciTech Connect (OSTI)

    Larsson, I.; Sihver, L.; Grundin, A.; Helmersson, J. O.

    2012-07-01

    To certify a continuous and safe operation of a boiling water reactor, careful surveillance of fuel integrity is of high importance. The detection of fuel failures can be performed by off-line gamma spectroscopy of off-gas samples and/or by on-line nuclide specific monitoring of gamma emitting noble gases. To establish the location of a leaking fuel rod, power suppression testing can be used. The accuracy of power suppression testing is dependent on the information of the delay time and the spreading of the released fission gases through the systems before reaching the sampling point. This paper presents a method to improve the accuracy of power suppression testing by determining the delay time and gas spreading profile. To estimate the delay time and examine the spreading of the gas in case of a fuel failure, helium was injected in the feed water system at Forsmark 3 nuclear power plant. The measurements were performed by using a helium detector system based on a mass spectrometer installed in the off-gas system. The helium detection system and the results of the experiment are presented in this paper. (authors)

  1. THE VALUE OF HELIUM-COOLED REACTOR TECHNOLOGIES OF NUCLEAR WASTE

    SciTech Connect (OSTI)

    C. RODRIGUEZ; A. BAXTER

    2001-03-01

    Helium-cooled reactor technologies offer significant advantages in accomplishing the waste transmutation process. They are ideally suited for use with thermal, epithermal, or fast neutron energy spectra. They can provide a relatively hard thermal neutron spectrum for transmutation of fissionable materials such as Pu-239 using ceramic-coated transmutation fuel particles, a graphite moderator, and a non-fertile burnable poison. These features (1) allow deep levels of transmutation with minimal or no intermediate reprocessing, (2) enhance passive decay heat removal via heat conduction and radiation, (3) allow operation at relatively high temperatures for a highly efficient generation of electricity, and (4) discharge the transmuted waste in a form that is highly resistant to corrosion for long times. They also offer the possibility for the use of epithermal neutrons that can interact with transmutable materials more effectively because of the large atomic cross sections in this energy domain. A fast spectrum may be useful for deep burnup of certain minor actinides. For this application, helium is essentially transparent to neutrons, does not degrade neutron energies, and offers the hardest possible neutron energy environment. In this paper, we report results from recent work on materials transmutation balances, safety, value to a geological repository, and economic considerations.

  2. Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights September 2010

    SciTech Connect (OSTI)

    Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D; Bell, Gary L

    2010-10-01

    The DB Program monthly highlights report for August 2010, ORNL/TM-2010/184, was distributed to program participants by email on September 17. This report discusses: (1) Core and Fuel Analysis - (a) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Prismatic Design (Logos), (b) Core Design Optimization in the HTR Pebble Bed Design (INL), (c) Microfuel analysis for the DB HTR (INL, GA, Logos); (2) Spent Fuel Management - (a) TRISO (tri-structural isotropic) repository behavior (UNLV), (b) Repository performance of TRISO fuel (UCB); (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor) - Synergy with other reactor fuel cycles (GA, Logos); (4) TRU (transuranic elements) HTR Fuel Qualification - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle - (a) Graphite Recycle (ORNL), (b) Aqueous Reprocessing, (c) Pyrochemical Reprocessing METROX (metal recovery from oxide fuel) Process Development (ANL).

  3. A passively-safe fusion reactor blanket with helium coolant and steel structure

    SciTech Connect (OSTI)

    Crosswait, K.M.

    1994-04-01

    Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accident conditions such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This thesis describes a new helium-cooled tritium breeding blanket concept which performs significantly better during such accidents than current designs. The proposed blanket uses reduced-activation ferritic steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m{sup 2}. The proposed geometry is based on the nested-shell concept developed by Wong, but some novel features are used to reduce the severity of the first wall temperature excursion. These features include the following: (1) A ``beryllium-joint`` concept is introduced, which allows solid beryllium slabs to be used as a thermal conduction path from the first wall to the cooler portions of the blanket. The joint concept allows for significant swelling of the beryllium (10 percent or more) without developing large stresses in the blanket structure. (2) Natural circulation of the coolant in the water-cooled shield is used to maintain shield temperatures below 100 degrees C, thus maintaining a heat sink close to the blanket during the accident. This ensures the long-term passive safety of the blanket.

  4. Double, Double Toil and Trouble: Tungsten Burns and Helium Bubbles...

    Office of Science (SC) [DOE]

    Helium bubbles are detrimental to plasma-facing materials such as tungsten in fusion reactors, which could serve as a possible new power source. Thus, understanding how helium ...

  5. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  6. Three-dimensional neutronics optimization of helium-cooled blanket for multi-functional experimental fusion-fission hybrid reactor (FDS-MFX)

    SciTech Connect (OSTI)

    Jiang, J.; Yuan, B.; Jin, M.; Wang, M.; Long, P.; Hu, L.

    2012-07-01

    Three-dimensional neutronics optimization calculations were performed to analyse the parameters of Tritium Breeding Ratio (TBR) and maximum average Power Density (PDmax) in a helium-cooled multi-functional experimental fusion-fission hybrid reactor named FDS (Fusion-Driven hybrid System)-MFX (Multi-Functional experimental) blanket. Three-stage tests will be carried out successively, in which the tritium breeding blanket, uranium-fueled blanket and spent-fuel-fueled blanket will be utilized respectively. In this contribution, the most significant and main goal of the FDS-MFX blanket is to achieve the PDmax of about 100 MW/m3 with self-sustaining tritium (TBR {>=} 1.05) based on the second-stage test with uranium-fueled blanket to check and validate the demonstrator reactor blanket relevant technologies based on the viable fusion and fission technologies. Four different enriched uranium materials were taken into account to evaluate PDmax in subcritical blanket: (i) natural uranium, (ii) 3.2% enriched uranium, (iii) 19.75% enriched uranium, and (iv) 64.4% enriched uranium carbide. These calculations and analyses were performed using a home-developed code VisualBUS and Hybrid Evaluated Nuclear Data Library (HENDL). The results showed that the performance of the blanket loaded with 64.4% enriched uranium was the most attractive and it could be promising to effectively obtain tritium self-sufficiency (TBR-1.05) and a high maximum average power density ({approx}100 MW/m{sup 3}) when the blanket was loaded with the mass of {sup 235}U about 1 ton. (authors)

  7. REACTOR

    DOE Patents [OSTI]

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  8. REACTOR

    DOE Patents [OSTI]

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  9. REACTOR

    DOE Patents [OSTI]

    Roman, W.G.

    1961-06-27

    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.

  10. REACTORS

    DOE Patents [OSTI]

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  11. ITER helium ash accumulation

    SciTech Connect (OSTI)

    Hogan, J.T.; Hillis, D.L.; Galambos, J.; Uckan, N.A. ); Dippel, K.H.; Finken, K.H. . Inst. fuer Plasmaphysik); Hulse, R.A.; Budny, R.V. . Plasma Physics Lab.)

    1990-01-01

    Many studies have shown the importance of the ratio {upsilon}{sub He}/{upsilon}{sub E} in determining the level of He ash accumulation in future reactor systems. Results of the first tokamak He removal experiments have been analysed, and a first estimate of the ratio {upsilon}{sub He}/{upsilon}{sub E} to be expected for future reactor systems has been made. The experiments were carried out for neutral beam heated plasmas in the TEXTOR tokamak, at KFA/Julich. Helium was injected both as a short puff and continuously, and subsequently extracted with the Advanced Limiter Test-II pump limiter. The rate at which the He density decays has been determined with absolutely calibrated charge exchange spectroscopy, and compared with theoretical models, using the Multiple Impurity Species Transport (MIST) code. An analysis of energy confinement has been made with PPPL TRANSP code, to distinguish beam from thermal confinement, especially for low density cases. The ALT-II pump limiter system is found to exhaust the He with maximum exhaust efficiency (8 pumps) of {approximately}8%. We find 1<{upsilon}{sub He}/{upsilon}{sub E}<3.3 for the database of cases analysed to date. Analysis with the ITER TETRA systems code shows that these values would be adequate to achieve the required He concentration with the present ITER divertor He extraction system.

  12. NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1959-10-27

    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.

  13. Neutronic reactor thermal shield

    DOE Patents [OSTI]

    Wende, Charles W. J.

    1976-06-15

    1. The method of operating a water-cooled neutronic reactor having a graphite moderator which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40-60 volume percent of the mixture, in contact with the graphite moderator.

  14. Reactor

    DOE Patents [OSTI]

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  15. ISOTHERMAL AIR-INGRESS VALIDATION EXPERIMENTS

    SciTech Connect (OSTI)

    Chang H. Oh; Eung S. Kim

    2013-01-01

    Idaho National Laboratory has conducted airingress experiments as part of a campaign to validate computational fluid dynamics (CFD) calculations for very high-temperature gas-cooled reactor (VHTR) analysis. An isothermal test loop was designed to recreate exchange or stratified flow that occurs in the lower plenum of VHTR after a break in the primary loop allows helium to leak out and reactor building air to enter the reactor core. The experiment was designed to measure stratified flow in the inlet pipe connecting to the lower plenum of the General Atomics gas turbine–modular helium reactor (GT-MHR). Instead of helium and air, brine and sucrose were used as heavy fluids, and water was used as the lighter fluid to create, using scaling laws, the appropriate flow characteristics of the lower plenum immediately after depressurization. These results clearly indicate that stratified flow is established even for very small density differences. Corresponding CFD results were validated with the experimental data. A grid sensitivity study on CFD models was also performed using the Richardson extrapolation and the grid convergence index method for the numerical accuracy of CFD calculations. The calculated current speed showed very good agreement with the experimental data, indicating that current CFD methods are suitable for simulating density gradient stratified flow phenomena in an air-ingress accident.

  16. Metal tritides helium emission

    SciTech Connect (OSTI)

    Beavis, L.C.

    1980-02-01

    Over the past several years, we have been measuring the release of helium from metal tritides (primarily erbium tritide). We find that qualitatively all tritides of interest to us behave the same. When they are first formed, the helium is released at a low rate that appears to be related to the amount of surface area which has access to the outside of the material (either film or bulk). For example, erbium tritide films initially release about 0.3% of the helium generated. Most tritide films emit helium at about this rate initially. At some later time, which depends upon the amount of helium generated, the parent occluding element and the degree of tritium saturation of the dihydride phase the helium emission changes to a new mode in which it is released at approximately the rate at which it is generated (for example, we measure this value to be approx. = .31 He/Er for ErT/sub 1/./sub 9/ films). If erbium ditritide is saturated beyond 1.9 T/Er, the critical helium/metal ratio decreases. For example, in bulk powders ErT/sub 2/./sub 15/ reaches critical release concentration at approx. = 0.03. Moderate elevation of temperature above room temperature has little impact on the helium release rate. It appears that the process may have approx. = 2 kcal/mol activation energy. The first helium formed is well bound. As the tritide ages, the helium is found in higher energy sites. Similar but less extensive measurements on scandium, titanium, and zirconium tritides are also described. Finally, the thermal desorption of erbium tritides of various ages from 50 days to 3154 days is discussed. Significant helium is desorbed along with the tritium in all but the youngest samples during thermodesorption.

  17. Helium Energy | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Energy Jump to: navigation, search Name: Helium Energy Place: Spain Sector: Renewable Energy Product: Spain-based renewable energy development company. References: Helium Energy1...

  18. Final Technical Report for the Period September 2002 through September 2005; H2-MHR Pre-Conceptual Design Report: SI-Based Plant; H2-MHR Pre-Conceptual Design Report: HTE-Based Plant

    SciTech Connect (OSTI)

    M. Richards; A. Shenoy; L. Brown; R. Buckingham; E. Harvego; K. Peddicord; M. Reza; J. Coupey

    2006-04-19

    For electricity and hydrogen production, an advanced reactor technology receiving considerable international interest is a modular, passively-safe version of the high-temperature, gas-cooled reactor, known in the U.S. as the Modular Helium Reactor (MHR), which operates at a power level of 600 MW(t). For electricity production, the MHR operates with an outlet helium temperature of 850 C to drive a direct, Brayton-cycle power-conversion system with a thermal-to-electrical conversion efficiency of 48 percent. This concept is referred to as the Gas Turbine MHR (GT-MHR). For hydrogen production, both electricity and process heat from the MHR are used to produce hydrogen. This concept is referred to as the H2-MHR. This report provides pre-conceptual design descriptions of full-scale, nth-of-a-kind H2 MHR plants based on thermochemical water splitting using the Sulfur-Iodine process and High-Temperature Electrolysis.

  19. H2-MHR Pre-Conceptual Design Report: SI-Based Plant; HTE-Based Plant

    SciTech Connect (OSTI)

    Matt Richards; A.S. Shenoy; L.C. Brown; R.T. Buckingham; E.A. Harvego; K.L. Peddicord; S.M.M. Reza; J.P. Coupey

    2006-04-19

    Hydrogen and electricity are expected to dominate the world energy system in the long term. The world currently consumes about 50 million metric tons of hydrogen per year, with the bulk of it being consumed by the chemical and refining industries. The demand for hydrogen is expected to increase, especially if the U.S. and other countries shift their energy usage towards a hydrogen economy, with hydrogen consumed as an energy commodity by the transportation, residential, and commercial sectors. However, there is strong motivation to not use fossil fuels in the future as a feedstock for hydrogen production, because the greenhouse gas carbon dioxide is a byproduct and fossil fuel prices are expected to increase significantly. For electricity and hydrogen production, an advanced reactor technology receiving considerable international interest is a modular, passively-safe version of the high-temperature, gas-cooled reactor (HTGR), known in the U.S. as the Modular Helium Reactor (MHR), which operates at a power level of 600 MW(t). For electricity production, the MHR operates with an outlet helium temperature of 850 C to drive a direct, Brayton-cycle power-conversion system (PCS) with a thermal-to-electrical conversion efficiency of 48 percent. This concept is referred to as the Gas Turbine MHR (GT-MHR). For hydrogen production, the process heat from the MHR is used to produce hydrogen. This concept is referred to as the H2-MHR.

  20. NEUTRONIC REACTOR SYSTEM

    DOE Patents [OSTI]

    Daniels, F.

    1957-10-15

    Gas-cooled solid-moderator type reactors wherein the fissionable fuel and moderator materials are each in the form of solid pebbles, or discrete particles, and are substantially homogeneously mixed in the proper proportion and placed within the core of the reactor are described. The shape of these discrete particles must be such that voids are present between them when mixed together. Helium enters the bottom of the core and passes through the voids between the fuel and moderator particles to absorb the heat generated by the chain reaction. The hot helium gas is drawn off the top of the core and may be passed through a heat exchanger to produce steam.

  1. Is solid helium a supersolid?

    SciTech Connect (OSTI)

    Hallock, Robert

    2015-05-15

    Recent experiments suggest that helium-4 atoms can flow through an experimental cell filled with solid helium. But that incompletely understood flow is quite different from the reported superfluid-like motion that so excited physicists a decade ago.

  2. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-07-14

    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.

  3. Cavity morphology in a Ni based superalloy under heavy ion irradiation with hot pre-injected helium. II

    SciTech Connect (OSTI)

    Zhang, He; Yao, Zhongwen, E-mail: yaoz@me.queensu.ca; Daymond, Mark R. [Department of Mechanical and Materials Engineering, Queen's University Kingston, Ontario K7L 3N6 (Canada); Kirk, Marquis A. [Material Science Division, Argonne National Laboratory Argonne, Illinois 60439 (United States)

    2014-03-14

    In the current investigation, TEM in-situ heavy ion (1?MeV Kr{sup 2+}) irradiation with helium pre-injected at elevated temperature (400?C) was conducted to simulate in-reactor neutron irradiation induced damage in CANDU spacer material Inconel X-750, in an effort to understand the effects of helium on irradiation induced cavity microstructures. Three different quantities of helium, 400 appm, 1000 appm, and 5000 appm, were pre-injected directly into TEM foils at 400?C. The samples containing helium were then irradiated in-situ with 1?MeV Kr{sup 2+} at 400?C to a final dose of 5.4 dpa (displacement per atom). Cavities were formed from the helium injection solely and the cavity density and size increased with increasing helium dosage. In contrast to previous heavy ion irradiations with cold pre-injected helium, heterogeneous nucleation of cavities was observed. During the ensuing heavy ion irradiation, dynamical observation showed noticeable size increase in cavities which nucleated close to the grain boundaries. A bubble-void transformation was observed after Kr{sup 2+} irradiation to high dose (5.4?dpa) in samples containing 1000 appm and 5000 appm helium. Cavity distribution was found to be consistent with in-reactor neutron irradiation induced cavity microstructures. This implies that the distribution of helium is greatly dependent on the injection temperature, and helium pre-injection at high temperature is preferred for simulating the migration of the transmutation produced helium.

  4. Helium segregation on surfaces of plasma-exposed tungsten

    DOE PAGES-Beta [OSTI]

    Maroudas, Dimitrios; Blondel, Sophie; Hu, Lin; Hammond, Karl D.; Wirth, Brian D.

    2016-01-21

    Here we report a hierarchical multi-scale modeling study of implanted helium segregation on surfaces of tungsten, considered as a plasma facing component in nuclear fusion reactors. We employ a hierarchy of atomic-scale simulations based on a reliable interatomic interaction potential, including molecular-statics simulations to understand the origin of helium surface segregation, targeted molecular-dynamics (MD) simulations of near-surface cluster reactions, and large-scale MD simulations of implanted helium evolution in plasma-exposed tungsten. We find that small, mobile He-n (1 <= n <= 7) clusters in the near-surface region are attracted to the surface due to an elastic interaction force that provides themore » thermodynamic driving force for surface segregation. Elastic interaction force induces drift fluxes of these mobile Hen clusters, which increase substantially as the migrating clusters approach the surface, facilitating helium segregation on the surface. Moreover, the clusters' drift toward the surface enables cluster reactions, most importantly trap mutation, in the near-surface region at rates much higher than in the bulk material. Moreover, these near-surface cluster dynamics have significant effects on the surface morphology, near-surface defect structures, and the amount of helium retained in the material upon plasma exposure. We integrate the findings of such atomic-scale simulations into a properly parameterized and validated spatially dependent, continuum-scale reaction-diffusion cluster dynamics model, capable of predicting implanted helium evolution, surface segregation, and its near-surface effects in tungsten. This cluster-dynamics model sets the stage for development of fully atomistically informed coarse-grained models for computationally efficient simulation predictions of helium surface segregation, as well as helium retention and surface morphological evolution, toward optimal design of plasma facing components.« less

  5. Electron Bubbles in Liquid Helium

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Electron Bubbles in Liquid Helium and Quantum Mechanics Humphrey J. Maris Brown University September 16, 2015 4:00 p.m. An electron entering liquid helium forces open a cavity referred to as an electron bubble. These objects have been studied in many past experiments and appear to be well understood. However, experiments have revealed that in addition to these normal electron bubbles there are other negatively charged objects in liquid helium. Despite much effort the structure of these so-called

  6. ISOTHERMAL AIR INGRESS VALIDATION EXPERIMENTS

    SciTech Connect (OSTI)

    Chang H Oh; Eung S Kim

    2011-09-01

    Idaho National Laboratory carried out air ingress experiments as part of validating computational fluid dynamics (CFD) calculations. An isothermal test loop was designed and set to understand the stratified-flow phenomenon, which is important as the initial air flow into the lower plenum of the very high temperature gas cooled reactor (VHTR) when a large break loss-of-coolant accident occurs. The unique flow characteristics were focused on the VHTR air-ingress accident, in particular, the flow visualization of the stratified flow in the inlet pipe to the vessel lower plenum of the General Atomic’s Gas Turbine-Modular Helium Reactor (GT-MHR). Brine and sucrose were used as heavy fluids, and water was used to represent a light fluid, which mimics a counter current flow due to the density difference between the stimulant fluids. The density ratios were changed between 0.87 and 0.98. This experiment clearly showed that a stratified flow between simulant fluids was established even for very small density differences. The CFD calculations were compared with experimental data. A grid sensitivity study on CFD models was also performed using the Richardson extrapolation and the grid convergence index method for the numerical accuracy of CFD calculations . As a result, the calculated current speed showed very good agreement with the experimental data, indicating that the current CFD methods are suitable for predicting density gradient stratified flow phenomena in the air-ingress accident.

  7. Helium dilution refrigeration system

    DOE Patents [OSTI]

    Roach, P.R.; Gray, K.E.

    1988-09-13

    A helium dilution refrigeration system operable over a limited time period, and recyclable for a next period of operation is disclosed. The refrigeration system is compact with a self-contained pumping system and heaters for operation of the system. A mixing chamber contains [sup 3]He and [sup 4]He liquids which are precooled by a coupled container containing [sup 3]He liquid, enabling the phase separation of a [sup 3]He rich liquid phase from a dilute [sup 3]He-[sup 4]He liquid phase which leads to the final stage of a dilution cooling process for obtaining low temperatures. The mixing chamber and a still are coupled by a fluid line and are maintained at substantially the same level with the still cross sectional area being smaller than that of the mixing chamber. This configuration provides maximum cooling power and efficiency by the cooling period ending when the [sup 3]He liquid is depleted from the mixing chamber with the mixing chamber nearly empty of liquid helium, thus avoiding unnecessary and inefficient cooling of a large amount of the dilute [sup 3]He-[sup 4]He liquid phase. 2 figs.

  8. Helium dilution refrigeration system

    DOE Patents [OSTI]

    Roach, Patrick R.; Gray, Kenneth E.

    1988-01-01

    A helium dilution refrigeration system operable over a limited time period, and recyclable for a next period of operation. The refrigeration system is compact with a self-contained pumping system and heaters for operation of the system. A mixing chamber contains .sup.3 He and .sup.4 He liquids which are precooled by a coupled container containing .sup.3 He liquid, enabling the phase separation of a .sup.3 He rich liquid phase from a dilute .sup.3 He-.sup.4 He liquid phase which leads to the final stage of a dilution cooling process for obtaining low temperatures. The mixing chamber and a still are coupled by a fluid line and are maintained at substantially the same level with the still cross sectional area being smaller than that of the mixing chamber. This configuration provides maximum cooling power and efficiency by the cooling period ending when the .sup.3 He liquid is depleted from the mixing chamber with the mixing chamber nearly empty of liquid helium, thus avoiding unnecessary and inefficient cooling of a large amount of the dilute .sup.3 He-.sup.4 He liquid phase.

  9. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Daniels, F.

    1962-12-18

    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)

  10. Pulsed helium ionization detection system

    DOE Patents [OSTI]

    Ramsey, Roswitha S. (Knoxville, TN); Todd, Richard A. (Knoxville, TN)

    1987-01-01

    A helium ionization detection system is provided which produces stable operation of a conventional helium ionization detector while providing improved sensitivity and linearity. Stability is improved by applying pulsed dc supply voltage across the ionization detector, thereby modifying the sampling of the detectors output current. A unique pulse generator is used to supply pulsed dc to the detector which has variable width and interval adjust features that allows up to 500 V to be applied in pulse widths ranging from about 150 nsec to about dc conditions.

  11. Pulsed helium ionization detection system

    DOE Patents [OSTI]

    Ramsey, R.S.; Todd, R.A.

    1985-04-09

    A helium ionization detection system is provided which produces stable operation of a conventional helium ionization detector while providing improved sensitivity and linearity. Stability is improved by applying pulsed dc supply voltage across the ionization detector, thereby modifying the sampling of the detectors output current. A unique pulse generator is used to supply pulsed dc to the detector which has variable width and interval adjust features that allows up to 500 V to be applied in pulse widths ranging from about 150 nsec to about dc conditions.

  12. Use of Multiple Reheat Helium Brayton Cycles to Eliminate the Intermediate Heat Transfer Loop for Advanced Loop Type SFRs

    SciTech Connect (OSTI)

    Haihua Zhao; Hongbin Zhang; Samuel E. Bays

    2009-05-01

    The sodium intermediate heat transfer loop is used in existing sodium cooled fast reactor (SFR) plant design as a necessary safety measure to separate the radioactive primary loop sodium from the water of the steam Rankine power cycle. However, the intermediate heat transfer loop significantly increases the SFR plant cost and decreases the plant reliability due to the relatively high possibility of sodium leakage. A previous study shows that helium Brayton cycles with multiple reheat and intercooling for SFRs with reactor outlet temperature in the range of 510°C to 650°C can achieve thermal efficiencies comparable to or higher than steam cycles or recently proposed supercritical CO2 cycles. Use of inert helium as the power conversion working fluid provides major advantages over steam or CO2 by removing the requirement for safety systems to prevent and mitigate the sodium-water or sodium-CO2 reactions. A helium Brayton cycle power conversion system therefore makes the elimination of the intermediate heat transfer loop possible. This paper presents a pre-conceptual design of multiple reheat helium Brayton cycle for an advanced loop type SFR. This design widely refers the new horizontal shaft distributed PBMR helium power conversion design features. For a loop type SFR with reactor outlet temperature 550°C, the design achieves 42.4% thermal efficiency with favorable power density comparing with high temperature gas cooled reactors.

  13. Assessment of Embrittlement of VHTR Structural Alloys in Impure Helium Environments

    SciTech Connect (OSTI)

    Crone, Wendy; Cao, Guoping; Sridhara, Kumar

    2013-05-31

    The helium coolant in high-temperature reactors inevitably contains low levels of impurities during steady-state operation, primarily consisting of small amounts of H{sub 2}, H{sub 2}O, CH{sub 4}, CO, CO{sub 2}, and N{sub 2} from a variety of sources in the reactor circuit. These impurities are problematic because they can cause significant long-term corrosion in the structural alloys used in the heat exchangers at elevated temperatures. Currently, the primary candidate materials for intermediate heat exchangers are Alloy 617, Haynes 230, Alloy 800H, and Hastelloy X. This project will evaluate the role of impurities in helium coolant on the stress-assisted grain boundary oxidation and creep crack growth in candidate alloys at elevated temperatures. The project team will: • Evaluate stress-assisted grain boundary oxidation and creep crack initiation and crack growth in the temperature range of 500-850°C in a prototypical helium environment. • Evaluate the effects of oxygen partial pressure on stress-assisted grain boundary oxidation and creep crack growth in impure helium at 500°C, 700°C, and 850°C respectively. • Characterize the microstructure of candidate alloys after long-term exposure to an impure helium environment in order to understand the correlation between stress-assisted grain boundary oxidation, creep crack growth, material composition, and impurities in the helium coolant. • Evaluate grain boundary engineering as a method to mitigate stress-assisted grain boundary oxidation and creep crack growth of candidate alloys in impure helium. The maximum primary helium coolant temperature in the high-temperature reactor is expected to be 850-1,000°C.Corrosion may involve oxidation, carburization, or decarburization mechanisms depending on the temperature, oxygen partial pressure, carbon activity, and alloy composition. These corrosion reactions can substantially affect long-term mechanical properties such as crack- growth rate and fracture

  14. Cryogenic system with GM cryocooler for krypton, xenon separation from hydrogen-helium purge gas

    SciTech Connect (OSTI)

    Chu, X. X.; Zhang, D. X.; Qian, Y.; Liu, W.; Zhang, M. M.; Xu, D.

    2014-01-29

    In the thorium molten salt reactor (TMSR), fission products such as krypton, xenon and tritium will be produced continuously in the process of nuclear fission reaction. A cryogenic system with a two stage GM cryocooler was designed to separate Kr, Xe, and H{sub 2} from helium purge gas. The temperatures of two stage heat exchanger condensation tanks were maintained at about 38 K and 4.5 K, respectively. The main fluid parameters of heat transfer were confirmed, and the structural heat exchanger equipment and cold box were designed. Designed concentrations after cryogenic separation of Kr, Xe and H{sub 2} in helium recycle gas are less than 1 ppb.

  15. Energy, helium, and the future: II

    SciTech Connect (OSTI)

    Krupka, M.C.; Hammel, E.F.

    1980-01-01

    The importance of helium as a critical resource material has been recognized specifically by the scientific community and more generally by the 1960 Congressional mandate to institute a long-range conservation program. A major study mandated by the Energy Reorganization Act of 1974 resulted in the publication in 1975 of the document, The Energy-Related Applications of Helium, ERDA-13. This document contained a comprehensive review and analysis relating to helium resources and present and future supply/demand relationships with particular emphasis upon those helium-dependent energy-related technologies projected to be implemented in the post-2000 year time period, e.g., fusion. An updated overview of the helium situation as it exists today is presented. Since publication of ERDA-13, important changes in the data base underlying that document have occurred. The data have since been reexamined, revised, and new information included. Potential supplies of helium from both conventional and unconventional natural gas resources, projected supply/demand relationships to the year 2030 based upon a given power-generation scenario, projected helium demand for specific energy-related technologies, and the supply options (national and international) available to meet that demand are discussed. An updated review will be given of the energy requirements for the extraction of helium from natural gas as they relate to the concentration of helium. A discussion is given concerning the technical and economic feasibility of several methods available both now and conceptually possible, to extract helium from helium-lean natural gas, the atmosphere, and outer space. Finally, a brief review is given of the 1980 Congressional activities with respect to the introduction and possible passage of new helium conservation legislation.

  16. The Next Generation Nuclear Plant - Insights Gained from the INEEL Point Design Studies

    SciTech Connect (OSTI)

    Philip E. MacDonald; A. M. Baxter; P. D. Bayless; J. M. Bolin; H. D. Gougar; R. L. Moore; A. M. Ougouag; M. B. Richards; R. L. Sant; J. W. Sterbentz; W. K. Terry

    2004-08-01

    This paper provides the results of an assessment of two possible versions of the Next Generation Nuclear Plant (NGNP), a prismatic fuel type helium gas-cooled reactor and a pebble-bed fuel helium gas reactor. Insights gained regarding the strengths and weaknesses of the two designs are also discussed. Both designs will meet the three basic requirements that have been set for the NGNP: a coolant outlet temperature of 1000 C, passive safety, and a total power output consistent with that expected for commercial high-temperature gas-cooled reactors. Two major modifications of the current Gas Turbine- Modular Helium Reactor (GT-MHR) design were needed to obtain a prismatic block design with a 1000 C outlet temperature: reducing the bypass flow and better controlling the inlet coolant flow distribution to the core. The total power that could be obtained for different core heights without exceeding a peak transient fuel temperature of 1600 C during a high or low-pressure conduction cooldown event was calculated. With a coolant inlet temperature of 490 C and 10% nominal core bypass flow, it is estimated that the peak power for a 10-block high core is 686 MWt, for a 12-block high core is 786 MWt, and for a 14-block core is about 889 MWt. The core neutronics calculations showed that the NGNP will exhibit strongly negative Doppler and isothermal temperature coefficients of reactivity over the burnup cycle. In the event of rapid loss of the helium gas, there is negligible core reactivity change. However, water or steam ingress into the core coolant channels can produce a relatively large reactivity effect. Two versions of an annular pebble-bed NGNP have also been developed, a 300 and a 600 MWt module. From this work we learned how to design passively safe pebble bed reactors that produce more than 600 MWt. We also found a way to improve both the fuel utilization and safety by modifying the pebble design (by adjusting the fuel zone radius in the pebble to optimize the fuel

  17. Surprisingly Large Generation and Retention of Helium and Hydrogen...

    Office of Scientific and Technical Information (OSTI)

    Journal Article: Surprisingly Large Generation and Retention of Helium and Hydrogen in ... Title: Surprisingly Large Generation and Retention of Helium and Hydrogen in Pure Nickel ...

  18. Helium Migration Mechanisms in Polycrystalline Uranium Dioxide

    SciTech Connect (OSTI)

    Martin, Guillaume; Desgardin, Pierre; Sauvage, Thierry; Barthe, Marie-France; Garcia, Philippe; Carlot, Gaelle

    2007-07-01

    This study aims at identifying the release mechanisms of helium in uranium dioxide. Two sets of polycrystalline UO{sub 2} sintered samples presenting different microstructures were implanted with {sup 3}He ions at concentrations in the region of 0.1 at.%. Changes in helium concentrations were monitored using two Nuclear Reaction Analysis (NRA) techniques based on the {sup 3}He(d,{alpha}){sup 1}H reaction. {sup 3}He release is measured in-situ during sample annealing at temperatures ranging between 700 deg. C and 1000 deg. C. Accurate helium depth profiles are generated after each annealing stage. Results that provide data for further understanding helium release mechanisms are discussed. It is found that helium diffusion appears to be enhanced above 900 deg. C in the vicinity of grain boundaries possibly as a result of the presence of defects. (authors)

  19. Helium refrigeration considerations for cryomodule design

    SciTech Connect (OSTI)

    Ganni, V.; Knudsen, P.

    2014-01-29

    Many of the present day accelerators are based on superconducting radio frequency (SRF) cavities, packaged in cryo-modules (CM), which depend on helium refrigeration at sub-atmospheric pressures, nominally 2 K. These specialized helium refrigeration systems are quite cost intensive to produce and operate. Particularly as there is typically no work extraction below the 4.5-K supply, it is important that the exergy loss between this temperature level and the CM load temperature(s) be minimized by the process configuration choices. This paper will present, compare and discuss several possible helium distribution process arrangements to support the CM loads.

  20. Tritium-helium effects in metals

    SciTech Connect (OSTI)

    Caskey, G.R.

    1985-09-01

    Investigations of helium effects in metals at the Savannah River Laboratory have been carried out by introducing helium by radioactive decay of tritium. This process does not create concurrent radiation damage, such as accompanies ion implantation and (n,..cap alpha..) reactions. The process has its own peculiarities, however, which partially mask and interact with the helium effect of interest. The distribution and local concentration of helium and tritium, which are responsible for changes in mechanical properties and fracture mode, are controlled by the large difference in solubility and diffusivity between the two atoms and by their differing interaction energies with lattice defects, impurities, and internal boundaries. Furthermore, in all investigations with helium generated from tritium decay, some tritium and deuterium are always present. Consequently, property changes include tritium-helium interaction effects to some extent. Results of investigations with several austenitic stainless steels, Armco iron, and niobium single crystals illustrate the variety of phenomena and some of the complex interactions that can be encountered.

  1. Tritium-helium effects in metals

    SciTech Connect (OSTI)

    Caskey, G.R. Jr.

    1985-01-01

    Investigations of helium effects in metals at the Savannah River Laboratory have been carried out by introducing helium by radioactive decay of tritium. This process does not create concurrent radiation damage, such as accompanies ion implantation and (n,..cap alpha..) reactions. The process has its own peculiarities, however, which partially mask and interact with the helium effect of interest. The distribution and local concentration of helium and tritium, which are responsible for changes in mechanical properties and fracture mode, are controlled by the large difference in solubility and diffusivity between the two atoms and by their differing interaction energies with lattice defects, impurities, and internal boundaries. Furthermore, in all investigations with helium generated from tritium decay, some tritium and deuterium are always present. Consequently, property changes include tritium-helium interaction effects to some extent. Results of investigations with several austenitic stainless steels, Armco iron, and niobium single crystals illustrate the variety of phenomena and some of the complex interactions that can be encountered.

  2. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  3. Helium, Iron and Electron Particle Transport and Energy Transport Studies on the TFTR Tokamak

    DOE R&D Accomplishments [OSTI]

    Synakowski, E. J.; Efthimion, P. C.; Rewoldt, G.; Stratton, B. C.; Tang, W. M.; Grek, B.; Hill, K. W.; Hulse, R. A.; Johnson, D .W.; Mansfield, D. K.; McCune, D.; Mikkelsen, D. R.; Park, H. K.; Ramsey, A. T.; Redi, M. H.; Scott, S. D.; Taylor, G.; Timberlake, J.; Zarnstorff, M. C. (Princeton Univ., NJ (United States). Plasma Physics Lab.); Kissick, M. W. (Wisconsin Univ., Madison, WI (United States))

    1993-03-01

    Results from helium, iron, and electron transport on TFTR in L-mode and Supershot deuterium plasmas with the same toroidal field, plasma current, and neutral beam heating power are presented. They are compared to results from thermal transport analysis based on power balance. Particle diffusivities and thermal conductivities are radially hollow and larger than neoclassical values, except possibly near the magnetic axis. The ion channel dominates over the electron channel in both particle and thermal diffusion. A peaked helium profile, supported by inward convection that is stronger than predicted by neoclassical theory, is measured in the Supershot The helium profile shape is consistent with predictions from quasilinear electrostatic drift-wave theory. While the perturbative particle diffusion coefficients of all three species are similar in the Supershot, differences are found in the L-Mode. Quasilinear theory calculations of the ratios of impurity diffusivities are in good accord with measurements. Theory estimates indicate that the ion heat flux should be larger than the electron heat flux, consistent with power balance analysis. However, theoretical values of the ratio of the ion to electron heat flux can be more than a factor of three larger than experimental values. A correlation between helium diffusion and ion thermal transport is observed and has favorable implications for sustained ignition of a tokamak fusion reactor.

  4. Lessons Learned From Gen I Carbon Dioxide Cooled Reactors

    SciTech Connect (OSTI)

    David E. Shropshire

    2004-04-01

    This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21st century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors.

  5. TRANSPARENT HELIUM IN STRIPPED ENVELOPE SUPERNOVAE

    SciTech Connect (OSTI)

    Piro, Anthony L.; Morozova, Viktoriya S., E-mail: piro@caltech.edu [Theoretical Astrophysics, California Institute of Technology, 1200 E. California Blvd., M/C 350-17, Pasadena, CA 91125 (United States)

    2014-09-01

    Using simple arguments based on photometric light curves and velocity evolution, we propose that some stripped envelope supernovae (SNe) show signs that a significant fraction of their helium is effectively transparent. The main pieces of evidence are the relatively low velocities with little velocity evolution, as are expected deep inside an exploding star, along with temperatures that are too low to ionize helium. This means that the helium should not contribute to the shaping of the main SN light curve, and thus the total helium mass may be difficult to measure from simple light curve modeling. Conversely, such modeling may be more useful for constraining the mass of the carbon/oxygen core of the SN progenitor. Other stripped envelope SNe show higher velocities and larger velocity gradients, which require an additional opacity source (perhaps the mixing of heavier elements or radioactive nickel) to prevent the helium from being transparent. We discuss ways in which similar analysis can provide insights into the differences and similarities between SNe Ib and Ic, which will lead to a better understanding of their respective formation mechanisms.

  6. SCREW COMPRESSOR CHARACTERISTICS FOR HELIUM REFRIGERATION SYSTEMS

    SciTech Connect (OSTI)

    Ganni, Venkatarao; Knudsen, Peter; Creel, Jonathan; Arenius, Dana; Casagrande, Fabio; Howell, Matt

    2008-03-01

    The oil injected screw compressors have practically replaced all other types of compressors in modern helium refrigeration systems due to their large displacement capacity, minimal vibration, reliability and capability of handling helium's high heat of compression.At the present state of compressor system designs for helium systems, typically two-thirds of the lost input power is due to the compression system. Therefore it is important to understand the isothermal and volumetric efficiencies of these machines to help properly design these compression systems to match the refrigeration process. This presentation summarizes separate tests that have been conducted on Sullair compressors at the Superconducting Super-Collider Laboratory (SSCL) in 1993, Howden compressors at Jefferson Lab (JLab) in 2006 and Howden compressors at the Spallation Neutron Source (SNS) in 2006. This work is part of an ongoing study at JLab to understand the theoretical basis for these efficiencies and their loss

  7. Cavity morphology in a Ni based superalloy under heavy ion irradiation with cold pre-injected helium. I

    SciTech Connect (OSTI)

    Zhang, He; Yao, Zhongwen Daymond, Mark R.; Kirk, Marquis A.

    2014-03-14

    In order to understand radiation damage in the nickel based superalloy Inconel X-750 in thermal reactors, where (n, ?) transmutation reaction also occurred in addition to fast neutron induced atomic displacement, heavy ion (1?MeV Kr{sup 2+}) irradiation with pre-injected helium was performed under in-situ observations of an intermediate voltage electron microscope at Argonne National Laboratory. By comparing to our previous studies using 1?MeV Kr{sup 2+} irradiation solely, the pre-injected helium was found to be essential in cavity nucleation. Cavities started to be visible after Kr{sup 2+} irradiation to 2.7 dpa at ?200?C in samples containing 200 appm, 1000 appm, and 5000 appm helium, respectively, but not at lower temperatures. The cavity growth was observed during the continuous irradiation. Cavity formation appeared along with a reduced number density of stacking fault tetrahedra, vacancy type defects. With higher pre-injected helium amount, a higher density of smaller cavities was observed. This is considered to be the result of local trapping effect of helium which disperses vacancies. The average cavity size increases with increasing irradiation temperatures; the density reduced; and the distribution of cavities became heterogeneous at elevated temperatures. In contrast to previous characterization of in-reactor neutron irradiated Inconel X-750, no obvious cavity sink to grain boundaries and phase boundaries was found even at high doses and elevated temperatures. MC-type carbides were observed as strong sources for agglomeration of cavities due to their enhanced trapping strength of helium and vacancies.

  8. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    SciTech Connect (OSTI)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  9. Helium damage in austenitic stainless steels

    SciTech Connect (OSTI)

    Caskey, G.R. Jr.; Mezzanotte, D.A. Jr.; Rawl, D.E. Jr.

    1983-01-01

    Helium produced by tritium decay was first shown to embrittle austenitic stainless steel at ambient temperature in tensile specimens of Nitronic-40 steel (Armco, Inc.). A long-term study was initiated to study this form of helium damage in five austenitic alloys. Results from this study have been analyzed by the J-integral technique and show a decrease in ductile fracture toughness with increasing He-3 concentration. Sustained-load cracking tests indicate that the stress intensity required to initiate and propagate a crack also decreases with increasing He-3 concentration. 9 figures, 3 tables.

  10. BOILING REACTORS

    DOE Patents [OSTI]

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  11. Transient analysis of an FHR coupled to a helium Brayton power cycle

    SciTech Connect (OSTI)

    Chen, Minghui; Kim, In Hun; Sun, Xiaodong; Christensen, Richard; Utgikar, Vivek; Sabharwall, Piyush

    2015-08-01

    The Fluoride salt-cooled High-temperature Reactor (FHR) features a passive decay heat removal system and a high-efficiency Brayton cycle for electricity generation. It typically employs an intermediate loop, consisting of an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX), to couple the primary system with the power conversion unit (PCU). In this study, a preliminary dynamic system model is developed to simulate transient characteristics of a prototypic 20-MWth Fluoride salt-cooled High-temperature Test Reactor (FHTR). The model consists of a series of differential conservation equations that are numerically solved using the MATLAB platform. For the reactor, a point neutron kinetics model is adopted. For the IHX and SHX, a fluted tube heat exchanger and an offset strip-fin heat exchanger are selected, respectively. Detailed geometric parameters of each component in the FHTR are determined based on the FHTR nominal steady-state operating conditions. Three initiating events are simulated in this study, including a positive reactivity insertion, a step increase in the mass flow rate of the PCU helium flow, and a step increase in the PCU helium inlet temperature to the SHX. The simulation results show that the reactor has inherent safety features for those three simulated scenarios. It is observed that the increase in the temperatures of the fuel pebbles and primary coolant is mitigated by the decrease in the reactor power due to negative temperature feedbacks. The results also indicate that the intermediate loop with the two heat exchangers plays a significant role in the transient progression of the integral reactor system.

  12. Continuum-scale Modeling of Hydrogen and Helium Bubble Growth...

    Office of Environmental Management (EM)

    Continuum-scale Modeling of Hydrogen and Helium Bubble Growth in Metals Continuum-scale Modeling of Hydrogen and Helium Bubble Growth in Metals Presentation from the 34th Tritium ...

  13. Fission-reactor experiments for fusion-materials research

    SciTech Connect (OSTI)

    Grossbeck, M.L.; Bloom, E.E.; Woods, J.W.; Vitek, J.M.; Thomas, K.R.

    1982-01-01

    The US Fusion Materials Program makes extensive use of fission reactors to study the effects of simulated fusion environments on materials and to develop improved alloys for fusion reactor service. The fast reactor, EBR-II, and the mixed spectrum reactors, HFIR and ORR, are all used in the fusion program. The HFIR and ORR produce helium from transmutations of nickel in a two-step thermal neutron absorption reaction beginning with /sup 58/Ni, and the fast neutrons in these reactors produce atomic displacements. The simultaneous effects of these phenomena produce damage similar to the very high energy neutrons of a fusion reactor. This paper describes irradiation capsules for mechanical property specimens used in the HFIR and the ORR. A neutron spectral tailoring experiment to achieve the fusion reactor He:dpa ratio will be discussed.

  14. Challenges in the Development of High Temperature Reactors

    SciTech Connect (OSTI)

    Piyush Sabharwall; Shannon M. Bragg-Sitton; Carl Stoots

    2013-10-01

    Advanced reactor designs offer potentially significant improvements over currently operating light water reactors including improved fuel utilization, increased efficiency, higher temperature operation (enabling a new suite of non-electric industrial process heat applications), and increased safety. As with most technologies, these potential performance improvements come with a variety of challenges to bringing advanced designs to the marketplace. There are technical challenges in material selection and thermal hydraulic and power conversion design that arise particularly for higher temperature, long life operation (possibly >60 years). The process of licensing a new reactor design is also daunting, requiring significant data collection for model verification and validation to provide confidence in safety margins associated with operating a new reactor design under normal and off-normal conditions. This paper focuses on the key technical challenges associated with two proposed advanced reactor concepts: the helium gas cooled Very High Temperature Reactor (VHTR) and the molten salt cooled Advanced High Temperature Reactor (AHTR).

  15. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Daniels, F.

    1959-10-27

    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.

  16. HEAVY WATER MODERATED NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Szilard, L.

    1958-04-29

    A nuclear reactor of the type which utilizes uranium fuel elements and a liquid coolant is described. The fuel elements are in the form of elongated tubes and are disposed within outer tubes extending through a tank containing heavy water, which acts as a moderator. The ends of the fuel tubes are connected by inlet and discharge headers, and liquid bismuth is circulated between the headers and through the fuel tubes for cooling. Helium is circulated through the annular space between the outer tubes in the tank and the fuel tubes to cool the water moderator to prevent boiling. The fuel tubes are covered with a steel lining, and suitable control means, heat exchange means, and pumping means for the coolants are provided to complete the reactor assembly.

  17. CONVECTION REACTOR

    DOE Patents [OSTI]

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

    1960-03-22

    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.

  18. NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1961-11-21

    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)

  19. Aberration-corrected X-ray spectrum imaging and Fresnel contrast to differentiate nanoclusters and cavities in helium-irradiated alloy 14YWT

    SciTech Connect (OSTI)

    Miller, Michael K; Parish, Chad M

    2014-01-01

    Helium accumulation negatively impacts structural materials used in neutron-irradiated environments, such as fission and fusion reactors. Next-generation fission and fusion reactors will require structural materials, such as steels, resistant to large neutron doses yet see service temperatures in the range most affected by helium embrittlement. Previous work has indicated the difficulty of experimentally differentiating nanometer-sized helium bubbles from the Ti-Y-O rich nanoclustsers (NCs) in radiation-tolerant nanostructured ferritic alloys (NFAs). Because the NCs are expected to sequester helium away from grain boundaries and reduce embrittlement, experimental methods to study simultaneously the NC and bubble populations are needed. In this study, aberration-corrected scanning transmission electron microscopy (STEM) results combining high-collection-efficiency X-ray spectrum images (SIs), multivariate statistical analysis (MVSA), and Fresnel-contrast bright-field STEM imaging have been used for such a purpose. Results indicate that Fresnel-contrast imaging, with careful attention to TEM-STEM reciprocity, differentiates bubbles from NCs, and MVSA of X-ray SIs unambiguously identifies NCs. Therefore, combined Fresnel-contrast STEM and X-ray SI is an effective STEM-based method to characterize helium-bearing NFAs.

  20. REACTOR COOLING

    DOE Patents [OSTI]

    Quackenbush, C.F.

    1959-09-29

    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.

  1. Characteristics of irradiation creep in the first wall of a fusion reactor

    SciTech Connect (OSTI)

    Coghlan, W.A.; Mansur, L.K.

    1981-01-01

    A number of significant differences in the irradiation environment of a fusion reactor are expected with respect to the fission reactor irradiation environment. These differences are expected to affect the characteristics of irradiation creep in the fusion reactor. Special conditions of importance are identified as the (1) large number of defects produced per pka, (2) high helium production rate, (3) cyclic operation, (4) unique stress histories, and (5) low temperature operations. Existing experimental data from the fission reactor environment is analyzed to shed light on irradiation creep under fusion conditions. Theoretical considerations are used to deduce additional characteristics of irradiation creep in the fusion reactor environment for which no experimental data are available.

  2. A novel scheme to handle highly pulsed loads with a standard helium refrigerator

    SciTech Connect (OSTI)

    Slack, D.S.

    1993-06-30

    Helium refrigerator performance degrades rapidly when it has to handle a varying or pulsed heat load. A novel scheme is presented to handle highly pulsed 4.5 K cryogenic loads with a standard helium refrigerator by isolating it from these pulses. The scheme uses a relatively simple arrangement of control valves, heat exchangers, and a storage dewar. Applications include pulsed tokamak machines such as TPX (Tokamak Physics Experiment) and ITER (International Thermonuclear Experimental Reactor). For example, the TPX (currently in the conceptual design phase in a DoE contract) requires an average 4.5 K refrigerator capacity of about 10 kW; however, pulsed loads caused by eddy current and nuclear heating will exceed 100 kW. The scheme presented here provides a method for handling these pulsed loads. Because of the simple and proven nature of the components involved and the thermodynamic properties of the helium, the system could be implemented for projects such as TPX or ITER with little or no development.

  3. THERMAL OSCILLATIONS IN LIQUID HELIUM TARGETS.

    SciTech Connect (OSTI)

    WANG,L.; JIA,L.X.

    2001-07-16

    A liquid helium target for the high-energy physics was built and installed in the proton beam line at the Alternate Gradient Synchrotron of Brookhaven National Laboratory in 2001. The target flask has a liquid volume of 8.25 liters and is made of thin Mylar film. A G-M/J-T cryocooler of five-watts at 4.2K was used to produce liquid helium and refrigerate the target. A thermosyphon circuit for the target was connected to the J-T circuit by a liquid/gas separator. Because of the large heat load to the target and its long transfer lines, thermal oscillations were observed during the system tests. To eliminate the oscillation, a series of tests and analyses were carried out. This paper describes the phenomena and provides the understanding of the thermal oscillations in the target system.

  4. Closed-loop pulsed helium ionization detector

    DOE Patents [OSTI]

    Ramsey, Roswitha S.; Todd, Richard A.

    1987-01-01

    A helium ionization detector for gas chromatography is operated in a constant current, pulse-modulated mode by configuring the detector, electrometer and a high voltage pulser in a closed-loop control system. The detector current is maintained at a fixed level by varying the frequency of fixed-width, high-voltage bias pulses applied to the detector. An output signal proportional to the pulse frequency is produced which is indicative of the charge collected for a detected species.

  5. Combined cold compressor/ejector helium refrigerator

    DOE Patents [OSTI]

    Brown, Donald P.

    1985-01-01

    A refrigeration apparatus having an ejector operatively connected with a cold compressor to form a two-stage pumping system. This pumping system is used to lower the pressure, and thereby the temperature of a bath of boiling refrigerant (helium). The apparatus as thus arranged and operated has substantially improved operating efficiency when compared to other processes or arrangements for achieving a similar low pressure.

  6. Combined cold compressor/ejector helium refrigerator

    DOE Patents [OSTI]

    Brown, D.P.

    1984-06-05

    A refrigeration apparatus having an ejector operatively connected with a cold compressor to form a two-stage pumping system. This pumping system is used to lower the pressure, and thereby the temperature of a bath of boiling refrigerant (helium). The apparatus as thus arranged and operated has substantially improved operating efficiency when compared to other processes or arrangements for achieving a similar low pressure.

  7. Preliminary issues associated with the next generation nuclear plant intermediate heat exchanger design.

    SciTech Connect (OSTI)

    Natesan, K.; Moisseytsev, A.; Majumdar, S.; Shankar, P. S.; Nuclear Engineering Division

    2007-04-05

    The Next Generation Nuclear Plant (NGNP), which is an advanced high temperature gas reactor (HTGR) concept with emphasis on production of both electricity and hydrogen, involves helium as the coolant and a closed-cycle gas turbine for power generation with a core outlet/gas turbine inlet temperature of 900-1000 C. In the indirect cycle system, an intermediate heat exchanger is used to transfer the heat from primary helium from the core to the secondary fluid, which can be helium, nitrogen/helium mixture, or a molten salt. The system concept for the vary high temperature reactor (VHTR) can be a reactor based on the prismatic block of the GT-MHR developed by a consortium led by General Atomics in the U.S. or based on the PBMR design developed by ESKOM of South Africa and British Nuclear Fuels of U.K. This report has made a preliminary assessment on the issues pertaining to the intermediate heat exchanger (IHX) for the NGNP. Two IHX designs namely, shell and tube and compact heat exchangers were considered in the assessment. Printed circuit heat exchanger, among various compact heat exchanger (HX) designs, was selected for the analysis. Irrespective of the design, the material considerations for the construction of the HX are essentially similar, except may be in the fabrication of the units. As a result, we have reviewed in detail the available information on material property data relevant for the construction of HX and made a preliminary assessment of several relevant factors to make a judicious selection of the material for the IHX. The assessment included four primary candidate alloys namely, Alloy 617 (UNS N06617), Alloy 230 (UNS N06230), Alloy 800H (UNS N08810), and Alloy X (UNS N06002) for the IHX. Some of the factors addressed in this report are the tensile, creep, fatigue, creep fatigue, toughness properties for the candidate alloys, thermal aging effects on the mechanical properties, American Society of Mechanical Engineers (ASME) Code compliance

  8. In situ controlled modification of the helium density in single helium-filled nanobubbles

    SciTech Connect (OSTI)

    David, M.-L. Pailloux, F.; Alix, K.; Mauchamp, V.; Pizzagalli, L.; Couillard, M.; Botton, G. A.

    2014-03-28

    We demonstrate that the helium density and corresponding pressure can be modified in single nano-scale bubbles embedded in semiconductors by using the electron beam of a scanning transmission electron microscope as a multifunctional probe: the measurement probe for imaging and chemical analysis and the irradiation source to modify concomitantly the pressure in a controllable way by fine tuning of the electron beam parameters. The control of the detrapping rate is achieved by varying the experimental conditions. The underlying physical mechanisms are discussed; our experimental observations suggest that the helium detrapping from bubbles could be interpreted in terms of direct ballistic collisions, leading to the ejection of the helium atoms from the bubble.

  9. Research questions reality of 'supersolid' in helium-4

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Research questions reality of 'supersolid' in helium-4 Research questions reality of 'supersolid' in helium-4 When cooled to temperatures below minus 452 degrees below zero Fahrenheit, helium-4 becomes a liquid-and an extraordinary liquid at that. May 17, 2011 Los Alamos National Laboratory sits on top of a once-remote mesa in northern New Mexico with the Jemez mountains as a backdrop to research and innovation covering multi-disciplines from bioscience, sustainable energy sources, to plasma

  10. NUCLEAR REACTOR

    DOE Patents [OSTI]

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

    1958-12-01

    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.

  11. Simplified Helium Refrigerator Cycle Analysis Using the `Carnot Step'

    SciTech Connect (OSTI)

    P. Knudsen; V. Ganni

    2006-05-01

    An analysis of the Claude form of an idealized helium liquefier for the minimum input work reveals the ''Carnot Step'' for helium refrigerator cycles. As the ''Carnot Step'' for a multi-stage polytropic compression process consists of equal pressure ratio stages; similarly for an idealized helium liquefier the ''Carnot Step'' consists of equal temperature ratio stages for a given number of expansion stages. This paper presents the analytical basis and some useful equations for the preliminary examination of existing and new Claude helium refrigeration cycles.

  12. Boron-10 Neutron Detectors for Helium-3 Replacement

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    efficiencies comparable to Helium-3 detectors, with demonstrated gamma neutron discrimination. Available for thumbnail of Feynman Center (505) 665-9090 Email Boron-10 Neutron...

  13. Helium isotopes in geothermal systems- Iceland, The Geysers,...

    Open Energy Information (Open El) [EERE & EIA]

    isotopes in geothermal systems- Iceland, The Geysers, Raft River and Steamboat Springs Jump to: navigation, search OpenEI Reference LibraryAdd to library Journal Article: Helium...

  14. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  15. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Wigner, E.P.

    1960-11-22

    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.

  16. REACTOR SHIELD

    DOE Patents [OSTI]

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

    1959-02-17

    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.

  17. Reactor apparatus

    DOE Patents [OSTI]

    Echtler, J. Paul

    1981-01-01

    A reactor apparatus for hydrocracking a polynuclear aromatic hydrocarbonaceous feedstock to produce lighter hydrocarbon fuels by contacting the hydrocarbonaceous feedstock with hydrogen in the presence of a molten metal halide catalyst.

  18. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  19. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  20. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  1. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  2. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

  3. Alternatives for Helium-3 in Multiplicity Counters

    SciTech Connect (OSTI)

    Ely, James H.; Siciliano, Edward R.; Lintereur, Azaree T.; Swinhoe, Martyn T.

    2013-04-01

    Alternatives to helium-3 are being actively pursued due to the shortage and rising costs of helium-3. For safeguards applications, there are a number of ongoing investigations to find alternatives that provide the same capability in a cost-effective manner. One of the greatest challenges is to find a comparable alternative for multiplicity counters, since they require high efficiency and short collection or die-away times. Work has been progressing on investigating three commercially available alternatives for high efficiency multiplicity counters: boron trifluoride (BF3) filled proportional tubes, boron-lined proportional tubes, and lithium fluoride with zinc sulfide coated light guides. The baseline multiplicity counter used for the investigation is the Epithermal Neutron Multiplicity Counter with 121 helium-3 filled tubes at 10 atmosphere pressure, which is a significant capability to match. The primary tool for the investigation has been modeling and simulation using the Monte Carlo N-Particle eXtended (MCNPX) radiation transport program, with experiments to validate the models. To directly calculate the coincidence rates in boron-lined (and possibly other) detectors, the MCNPX code has been enhanced to allow the existing coincidence tally to be used with energy deposition rather than neutron capture reactions. This allows boron-lined detectors to be modeled more accurately. Variations of tube number and diameter along with variations in the amount of inter-tube moderator have been conducted for the BF3 and boron-lined cases. Tube pressure was investigated for BF3, up to two atmospheres, as well as optimal boron thickness in the boron-lined tubes. The lithium fluoride was modeled as sheets of material with light guides in between, and the number and thickness of the sheets investigated. The amount of light guide, which in this case doubles as a moderator, was also optimized. The results of these modeling and simulation optimization investigations are described

  4. Helium transport and ash control studies

    SciTech Connect (OSTI)

    Miley, G.H.

    1992-01-01

    The Primary goal of this research is to develop a helium (ash) transport scaling law based on experimental data from devices such as TFTR and JET. To illustrate the importance of this, we have studied ash accumulation effects on ignition requirements using a O-D transport model. Ash accumulation is characterized in the model by the ratio of the helium particle confinement time to the energy confinement time t{sub {alpha}}/t{sub E}. Results show that the ignition window'' shrinks rapidly as t{sub {alpha}}/t{sub E} increases, closing for high t{sub {alpha}}/t{sub E} increases, closing for high t{sub {alpha}}/t{sub E}. A best'' value for t{sub {alpha}}/t{sub E} will ultimately be determined from our scaling law studies. A helium transport scaling law is being sought that expresses the transport coefficients (D{sub {alpha}}, V{sub {alpha}}) as a function of the local plasma parameters. This is necessary for use in transport code calculations, e.g. for BALDUR. Based on experimental data from L-mode plasma operation in TFTR, a scaling law to a power law expression has been obtained using a least-square fit method. It is found that the transport coefficients are strongly affected by the local magnetic field and safety factor q. A preliminary conclusion from this work is that active control of ash buildup must be developed. To study control, we have developed a O-D plasma model which employs a simple pole-placement control model. Some preliminary calculations with this model are presented.

  5. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    SciTech Connect (OSTI)

    Hongbin Zhang; Haihua Zhao

    2009-05-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.

  6. Helium-3 and Helium-4 acceleration by high power laser pulses for hadron therapy

    SciTech Connect (OSTI)

    Bulanov, S. S.; Esarey, E.; Schroeder, C. B.; Leemans, W. P.; Bulanov, S. V.; Margarone, D.; Korn, G.; Haberer, T.

    2015-06-24

    The laser driven acceleration of ions is considered a promising candidate for an ion source for hadron therapy of oncological diseases. Though proton and carbon ion sources are conventionally used for therapy, other light ions can also be utilized. Whereas carbon ions require 400 MeV per nucleon to reach the same penetration depth as 250 MeV protons, helium ions require only 250 MeV per nucleon, which is the lowest energy per nucleon among the light ions. This fact along with the larger biological damage to cancer cells achieved by helium ions, than that by protons, makes this species an interesting candidate for the laser driven ion source. Two mechanisms (Magnetic Vortex Acceleration and hole-boring Radiation Pressure Acceleration) of PW-class laser driven ion acceleration from liquid and gaseous helium targets are studied with the goal of producing 250 MeV per nucleon helium ion beams that meet the hadron therapy requirements. We show that He3 ions, having almost the same penetration depth as He4 with the same energy per nucleon, require less laser power to be accelerated to the required energy for the hadron therapy.

  7. Production of thorium-229 using helium nuclei

    DOE Patents [OSTI]

    Mirzadeh, Saed (Knoxville, TN) [Knoxville, TN; Garland, Marc Alan (Knoxville, TN) [Knoxville, TN

    2010-12-14

    A method for producing .sup.229Th includes the steps of providing .sup.226Ra as a target material, and bombarding the target material with alpha particles, helium-3, or neutrons to form .sup.229Th. When neutrons are used, the neutrons preferably include an epithermal neutron flux of at least 1.times.10.sup.13 n s.sup.-1cm.sup.-2. .sup.228Ra can also be bombarded with thermal and/or energetic neutrons to result in a neutron capture reaction to form .sup.229Th. Using .sup.230Th as a target material, .sup.229Th can be formed using neutron, gamma ray, proton or deuteron bombardment.

  8. The primordial helium abundance from updated emissivities

    SciTech Connect (OSTI)

    Aver, Erik; Olive, Keith A.; Skillman, Evan D.; Porter, R.L. E-mail: olive@umn.edu E-mail: skillman@astro.umn.edu

    2013-11-01

    Observations of metal-poor extragalactic H II regions allow the determination of the primordial helium abundance, Y{sub p}. The He I emissivities are the foundation of the model of the H II region's emission. Porter, Ferland, Storey, and Detisch (2012) have recently published updated He I emissivities based on improved photoionization cross-sections. We incorporate these new atomic data and update our recent Markov Chain Monte Carlo analysis of the dataset published by Izotov, Thuan, and Stasi'nska (2007). As before, cuts are made to promote quality and reliability, and only solutions which fit the data within 95% confidence level are used to determine the primordial He abundance. The previously qualifying dataset is almost entirely retained and with strong concordance between the physical parameters. Overall, an upward bias from the new emissivities leads to a decrease in Y{sub p}. In addition, we find a general trend to larger uncertainties in individual objects (due to changes in the emissivities) and an increased variance (due to additional objects included). From a regression to zero metallicity, we determine Y{sub p} = 0.2465 ± 0.0097, in good agreement with the BBN result, Y{sub p} = 0.2485 ± 0.0002, based on the Planck determination of the baryon density. In the future, a better understanding of why a large fraction of spectra are not well fit by the model will be crucial to achieving an increase in the precision of the primordial helium abundance determination.

  9. NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1958-09-01

    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.

  10. Initial assessment of environmental effects on SiC/SiC composites in helium-cooled nuclear systems

    SciTech Connect (OSTI)

    Contescu, Cristian I

    2013-09-01

    This report summarized the information available in the literature on the chemical reactivity of SiC/SiC composites and of their components in contact with the helium coolant used in HTGR, VHTR and GFR designs. In normal operation conditions, ultra-high purity helium will have chemically controlled impurities (water, oxygen, carbon dioxide, carbon monoxide, methane, hydrogen) that will create a slightly oxidizing gas environment. Little is known from direct experiments on the reactivity of third generation (nuclear grade) SiC/SiC composites in contact with low concentrations of water or oxygen in inert gas, at high temperature. However, there is ample information about the oxidation in dry and moist air of SiC/SiC composites at high temperatures. This information is reviewed first in the next chapters. The emphasis is places on the improvement in material oxidation, thermal, and mechanical properties during three stages of development of SiC fibers and at least two stages of development of the fiber/matrix interphase. The chemical stability of SiC/SiC composites in contact with oxygen or steam at temperatures that may develop in off-normal reactor conditions supports the conclusion that most advanced composites (also known as nuclear grade SiC/SiC composites) have the chemical resistance that would allow them maintain mechanical properties at temperatures up to 1200 1300 oC in the extreme conditions of an air or water ingress accident scenario. Further research is needed to assess the long-term stability of advanced SiC/SiC composites in inert gas (helium) in presence of very low concentrations (traces) of water and oxygen at the temperatures of normal operation of helium-cooled reactors. Another aspect that needs to be investigated is the effect of fast neutron irradiation on the oxidation stability of advanced SiC/SiC composites in normal operation conditions.

  11. Double photoionization of helium with synchrotron x-rays: Proceedings

    SciTech Connect (OSTI)

    Not Available

    1994-01-01

    This report contains papers on the following topics: Overview and comparison of photoionization with charged particle impact; The ratio of double to single ionization of helium: the relationship of photon and bare charged particle impact ionization; Double photoionization of helium at high energies; Compton scattering of photons from electrons bound in light elements; Electron ionization and the Compton effect in double ionization of helium; Elimination of two atomic electrons by a single energy photon; Double photoionization of helium at intermediate energies; Double Photoionization: Gauge Dependence, Coulomb Explosion; Single and Double Ionization by high energy photon impact; The effect of Compton Scattering on the double to single ionization ratio in helium; and Double ionization of He by photoionization and Compton scattering. These papers have been cataloged separately for the database.

  12. Detailed and simplified nonequilibrium helium ionization in the solar atmosphere

    SciTech Connect (OSTI)

    Golding, Thomas Peter; Carlsson, Mats; Leenaarts, Jorrit E-mail: mats.carlsson@astro.uio.no

    2014-03-20

    Helium ionization plays an important role in the energy balance of the upper chromosphere and transition region. Helium spectral lines are also often used as diagnostics of these regions. We carry out one-dimensional radiation-hydrodynamics simulations of the solar atmosphere and find that the helium ionization is set mostly by photoionization and direct collisional ionization, counteracted by radiative recombination cascades. By introducing an additional recombination rate mimicking the recombination cascades, we construct a simplified three-level helium model atom consisting of only the ground states. This model atom is suitable for modeling nonequilibrium helium ionization in three-dimensional numerical models. We perform a brief investigation of the formation of the He I 10830 and He II 304 spectral lines. Both lines show nonequilibrium features that are not recovered with statistical equilibrium models, and caution should therefore be exercised when such models are used as a basis for interpretating observations.

  13. Catalytic reactor

    DOE Patents [OSTI]

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    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.

  14. POWER REACTOR

    DOE Patents [OSTI]

    Zinn, W.H.

    1958-07-01

    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.

  15. REACTOR CONTROL

    DOE Patents [OSTI]

    Fortescue, P.; Nicoll, D.

    1962-04-24

    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)

  16. Bioconversion reactor

    DOE Patents [OSTI]

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  17. NEUTRONIC REACTORS

    DOE Patents [OSTI]

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

    1958-10-14

    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.

  18. (Reactor dosimetry)

    SciTech Connect (OSTI)

    West, C.D.

    1990-09-13

    The lead in most aspects of research reactor design and use passed from the USA about 15 years ago, soon after the construction of the HFIR and HFBR. The Europeans have consistently upgraded and improved their existing facilities and have built new ones including the HFR at Grenoble and ORPHEE at Saclay. They studied ultra-high flux concepts ({approximately}10{sup 20}/m{sup {minus}2}{center dot}s{sup {minus}1}) about 10 years ago, and are in the design phase of a new, highly efficient medium flux reactor to be built at Garching, near Munich in Germany. A visit was made to Interatom, the firm -- the equivalent of the Architect/Engineer for the ANS project -- responsible, under contract to the Technical University of Munich, for the new Munich reactor design. There are many similarities to the ANS design, and we reviewed and discussed technical and safety aspects of the two reactors. A request was made for some new, hitherto proprietary, experimental data on reactor thermal hydraulics and cooling that will be very valuable to the ANS project. I presented a seminar on the ANS project. A visit was made to Kernforschungszentrum Karlsruhe and knowledge was gained from Dr. Kuchle, a true pioneer of ultra-high flux reactor concepts, of their work. Dr. Kuchle kindly reviewed the ANS reference core and cooling system design (with favorable conclusions). I then talked with researchers working on materials irradiation damage and activation of structural materials by neutron irradiation, both key issues for the ANS. I was shown some new techniques they have developed for testing materials irradiation effects at high fluences, in a short time, using accelerated particle beams.

  19. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Szilard, L.

    1957-09-24

    Reactors of the type employing plates of natural uranium in a moderator are discussed wherein the plates are um-formly disposed in parallel relationship to each other thereby separating the moderator material into distinct and individual layers. Each plate has an uninterrupted sunface area substantially equal to the cross-sectional area of the active portion of the reactor, the particular size of the plates and the volume ratio of moderator to uranium required to sustain a chain reaction being determinable from the known purity of these materials and other characteristics such as the predictable neutron losses due to the formation of radioactive elements of extremely high neutron capture cross section.

  20. NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1958-04-15

    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.

  1. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Young, G.

    1963-01-01

    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

  2. Steady-state tokamak reactor with non-divertor impurity control: STARFIRE

    SciTech Connect (OSTI)

    Baker, C.C.

    1980-01-01

    STARFIRE is a conceptual design study of a commercial tokamak fusion electric power plant. Particular emphasis has been placed on simplifying the reactor concept by developing design concepts to produce a steady-state tokamak with non-divertor impurity control and helium ash removal. The concepts of plasma current drive using lower hybrid rf waves and a limiter/vacuum system for reactor applications are described.

  3. Superconducting cable cooling system by helium gas and a mixture of gas and liquid helium

    DOE Patents [OSTI]

    Dean, John W.

    1977-01-01

    Thermally contacting, oppositely streaming cryogenic fluid streams in the same enclosure in a closed cycle that changes from a cool high pressure helium gas to a cooler reduced pressure helium fluid comprised of a mixture of gas and boiling liquid so as to be near the same temperature but at different pressures respectively in go and return legs that are in thermal contact with each other and in thermal contact with a longitudinally extending superconducting transmission line enclosed in the same cable enclosure that insulates the line from the ambient at a temperature T.sub.1. By first circulating the fluid in a go leg from a refrigerator at one end of the line as a high pressure helium gas near the normal boiling temperature of helium; then circulating the gas through an expander at the other end of the line where the gas becomes a mixture of reduced pressure gas and boiling liquid at its boiling temperature; then by circulating the mixture in a return leg that is separated from but in thermal contact with the gas in the go leg and in the same enclosure therewith; and finally returning the resulting low pressure gas to the refrigerator for compression into a high pressure gas at T.sub.2 is a closed cycle, where T.sub.1 >T.sub.2, the temperature distribution is such that the line temperature is nearly constant along its length from the refrigerator to the expander due to the boiling of the liquid in the mixture. A heat exchanger between the go and return lines removes the gas from the liquid in the return leg while cooling the go leg.

  4. Compact hydrogen/helium isotope mass spectrometer

    DOE Patents [OSTI]

    Funsten, Herbert O.; McComas, David J.; Scime, Earl E.

    1996-01-01

    The compact hydrogen and helium isotope mass spectrometer of the present invention combines low mass-resolution ion mass spectrometry and beam-foil interaction technology to unambiguously detect and quantify deuterium (D), tritium (T), hydrogen molecule (H.sub.2, HD, D.sub.2, HT, DT, and T.sub.2), .sup.3 He, and .sup.4 He concentrations and concentration variations. The spectrometer provides real-time, high sensitivity, and high accuracy measurements. Currently, no fieldable D or molecular speciation detectors exist. Furthermore, the present spectrometer has a significant advantage over traditional T detectors: no confusion of the measurements by other beta-emitters, and complete separation of atomic and molecular species of equivalent atomic mass (e.g., HD and .sup.3 He).

  5. Helium isotopes and tectonics in southern Italy

    SciTech Connect (OSTI)

    Sano, Yuji; Wakita, Hiroshi ); Nuccio, M.P. ); Italiano, F.

    1989-06-01

    Geodynamic evolution of southern Italy can be understood within the framework of the Mediterranean-Alpine System. Subduction of a plate along the Sicily-Calabrian forearc under the Tyrrhenian Sea has been suggested by many geophysicists, although it is not yet confirmed and remains somewhat controversial. Helium isotope ratios provide useful information on the geotectonic structure of the region. The authors report here the {sup 3}H/{sup 4}He ratios of terrestrial gas samples from southern Italy. The observed {sup 3}He/{sup 4}He ratios are relatively high in the Eolian volcanic arc region and low in the other areas. Dichotomous explanations are presented. Firstly, volcanic arc-forearc hypothesis suggests the subduction along the Sicily-Calabrian forearc. Secondly, horizontal transport hypothesis is described based on the relationship between the ratios and radial distance from the recent spreading basin in Southern Tyrrhenian Sea.

  6. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  7. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Starr, C.

    1963-01-01

    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)

  8. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J.

    1976-08-17

    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.

  9. Neutronic reactor

    DOE Patents [OSTI]

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  10. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Anderson, H.L.

    1958-10-01

    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.

  11. The modular high-temperature gas-cooled reactor (MHTGR)

    SciTech Connect (OSTI)

    Neylan, A.J.

    1986-10-01

    The MHTGR is an advanced reactor concept being developed in the USA under a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes basic HTGR features of ceramic fuel, helium coolant and a graphite moderator. However the specific size and configuration are selected to utilize the inherently safe characteristics associated with these standard features coupled with passive safety systems to provide a significantly higher margin of safety and investment protection than current generation reactors. Evacuation or sheltering of the public is not required. The major components of the nuclear steam supply, with special emphasis on the core, are described. Safety assessments of the concept are discussed.

  12. REACTOR MONITORING

    DOE Patents [OSTI]

    Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

    1959-02-01

    A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

  13. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Treshow, M.

    1958-08-19

    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.

  14. REACTOR UNLOADING

    DOE Patents [OSTI]

    Leverett, M.C.

    1958-02-18

    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.

  15. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wade, E.J.

    1958-09-16

    This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

  16. Nuclear reactor

    DOE Patents [OSTI]

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  17. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  18. NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  19. Neutronic reactor

    DOE Patents [OSTI]

    Lewis, Warren R.

    1978-05-30

    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.

  20. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  1. How to make Raman-inactive helium visible in Raman spectra of tritium-helium gas mixtures

    SciTech Connect (OSTI)

    Schloesser, M.; Pakari, O.; Rupp, S.; Mirz, S.; Fischer, S.

    2015-03-15

    Raman spectroscopy, a powerful method for the quantitative compositional analysis of molecular gases, e.g. mixtures of hydrogen isotopologues, is not able to detect monoatomic species like helium. This deficit can be overcome by using radioluminescence emission from helium atoms induced by β-electrons from tritium decay. We present theoretical considerations and combined Raman/radioluminescence spectra. Furthermore, we discuss the linearity of the method together with validation measurements for determining the pressure dependence. Finally, we conclude how this technique can be used for samples of helium with traces of tritium, and vice versa. (authors)

  2. REACTOR CONTROL

    DOE Patents [OSTI]

    Ruano, W.J.

    1957-12-10

    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.

  3. Technique to eliminate helium induced weld cracking in stainless steels

    SciTech Connect (OSTI)

    Chin-An Wang; Chin, B.A.; Grossbeck, M.L.

    1992-12-31

    Experiments have shown that Type 316 stainless steel is susceptible to heat-affected-zone (HAZ) cracking upon cooling when welded using the gas tungsten arc (GTA) process under lateral constraint. The cracking has been hypothesized to be caused by stress-assisted helium bubble growth and rupture at grain boundaries. This study utilized an experimental welding setup which enabled different compressive stresses to be applied to the plates during welding. Autogenous GTA welds were produced in Type 316 stainless steel doped with 256 appm helium. The application of a compressive stress, 55 Mpa, during welding suppressed the previously observed catastrophic cracking. Detailed examinations conducted after welding showed a dramatic change in helium bubble morphology. Grain boundary bubble growth along directions parallel to the weld was suppressed. Results suggest that stress-modified welding techniques may be used to suppress or eliminate helium-induced cracking during joining of irradiated materials.

  4. Ab initio study of helium behavior in titanium tritides

    SciTech Connect (OSTI)

    Liang, J. H.; Dai, Yunya; Yang, Li; Peng, SM; Fan, K. M.; Long, XG; Zhou, X. S.; Zu, Xiaotao; Gao, Fei

    2013-03-01

    Ab initio calculations based on density functional theory have been performed to investigate the relative stability of titanium tritides and the helium behavior in stable titanium tritides. The results show that the β-phase TiT1.5 without two tritium along the [100] direction (TiT1.5[100]) is more stable than other possible structures. The stability of titanium tritides decrease with the increased generation of helium in TiT1.5[100]. In addition, helium generated by tritium decay prefers locating at a tetrahedral site, and favorably migrates between two neighbor vacant tetrahedral sites through an intermediate octahedral site in titanium tritides, with a migration energy of 0.23 eV. Furthermore, helium is easily accumulated on a (100) plane in β-phase TiT1.5[100].

  5. Demonstrating Strong Electric Fields in Liquid Helium for Tests...

    Office of Science (SC) [DOE]

    Image courtesy of Los Alamos National Laboratory The Medium Scale High Voltage Test apparatus in TA-53 Building 10 allowed scientists to test electric fields in liquid helium, a ...

  6. Underground helium travels to the Earth's surface via aquifers...

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Tweet EmailPrint Before it can put the party in party balloons, helium is carried from deep within the Earth's crust to the surface via aquifers, according to new research...

  7. The Hall D solenoid helium refrigeration system at JLab

    SciTech Connect (OSTI)

    Laverdure, Nathaniel A.; Creel, Jonathan D.; Dixon, Kelly d.; Ganni, Venkatarao; Martin, Floyd D.; Norton, Robert O.; Radovic, Sasa

    2014-01-01

    Hall D, the new Jefferson Lab experimental facility built for the 12GeV upgrade, features a LASS 1.85 m bore solenoid magnet supported by a 4.5 K helium refrigerator system. This system consists of a CTI 2800 4.5 K refrigerator cold box, three 150 hp screw compressors, helium gas management and storage, and liquid helium and nitrogen storage for stand-alone operation. The magnet interfaces with the cryo refrigeration system through an LN2-shielded distribution box and transfer line system, both designed and fabricated by JLab. The distribution box uses a thermo siphon design to respectively cool four magnet coils and shields with liquid helium and nitrogen. We describe the salient design features of the cryo system and discuss our recent commissioning experience.

  8. Helium Pumping Wall for a Liquid Lithium Tokamak Richard Majeski |

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Princeton Plasma Physics Lab Helium Pumping Wall for a Liquid Lithium Tokamak Richard Majeski This invention is designed to be a subsystem of a device, a tokamak with walls or plasma facing components of liquid lithium. This approach to constructing the lithium-bearing walls of the tokamak allows the wall to fulfill a necessary function -- helium pumping - for which a complex structure was formerly required. The primary novel feature of the invention is that a permeable wall is used to

  9. Recovery of purified helium or hydrogen from gas mixtures

    DOE Patents [OSTI]

    Merriman, J.R.; Pashley, J.H.; Stephenson, M.J.; Dunthorn, D.I.

    1974-01-15

    A process is described for the removal of helium or hydrogen from gaseous mixtures also containing contaminants. The gaseous mixture is contacted with a liquid fluorocarbon in an absorption zone maintained at superatomspheric pressure to preferentially absorb the contaminants in the fluorocarbon. Unabsorbed gas enriched in hydrogen or helium is withdrawn from the absorption zone as product. Liquid fluorocarbon enriched in contaminants is withdrawn separately from the absorption zone. (10 claims)

  10. Process Options for Nominal 2-K Helium Refrigeration System Designs

    SciTech Connect (OSTI)

    Peter Knudsen, Venkatarao Ganni

    2012-07-01

    Nominal 2-K helium refrigeration systems are frequently used for superconducting radio frequency and magnet string technologies used in accelerators. This paper examines the trade-offs and approximate performance of four basic types of processes used for the refrigeration of these technologies; direct vacuum pumping on a helium bath, direct vacuum pumping using full or partial refrigeration recovery, cold compression, and hybrid compression (i.e., a blend of cold and warm sub-atmospheric compression).

  11. Generation and Retention of Helium and Hydrogen in Austenitic...

    Office of Scientific and Technical Information (OSTI)

    Steels Irradiated in a Variety of LWR and Test Reactor Spectral Environments Citation ... Steels Irradiated in a Variety of LWR and Test Reactor Spectral Environments In fission ...

  12. Generation and Retention of Helium and Hydrogen in Austenitic...

    Office of Scientific and Technical Information (OSTI)

    Steels Irradiated in a Variety of LWR and Test Reactor Spectral Environments Citation ... Steels Irradiated in a Variety of LWR and Test Reactor Spectral Environments You are ...

  13. Helium Loop Cooling Channel Hydraulic Characterization

    SciTech Connect (OSTI)

    Olivas, Eric Richard; Morgan, Robert Vaughn; Woloshun, Keith Albert

    2015-07-02

    New methods for generating ⁹⁹Mo are being explored in an effort to eliminate proliferation issues and provide a domestic supply of ⁹⁹mTc for medical imaging. Electron accelerating technology is used by sending an electron beam through a series of ¹⁰⁰Mo targets. During this process a large amount of heat is created, which directly affects the operating temperature set for the system. In order to maintain the required temperature range, helium gas is used to serve as a cooling agent that flows through narrow channels between the target disks. Currently we are tailoring the cooling channel entrance and exits to decrease the pressure drop through the targets. Currently all hardware has be procured and manufactured to conduct flow measurements and visualization via solid particle seeder. Pressure drop will be studied as a function of mass flow and diffuser angle. The results from these experiments will help in determining target cooling geometry and validate CFD code results.

  14. B Reactor | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Operational Management History Manhattan Project Signature Facilities B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first ...

  15. Nuclear reactor

    DOE Patents [OSTI]

    Pennell, William E.; Rowan, William J.

    1977-01-01

    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.

  16. NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Ashley, J.W.

    1958-12-16

    A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

  17. Photocatalytic reactor

    DOE Patents [OSTI]

    Bischoff, B.L.; Fain, D.E.; Stockdale, J.A.D.

    1999-01-19

    A photocatalytic reactor is described 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. 4 figs.

  18. Optimum Reactor Outlet Temperatures for High Temperature Gas-Cooled Reactors Integrated with Industrial Processes

    SciTech Connect (OSTI)

    Lee O. Nelson

    2011-04-01

    This report summarizes the results of a temperature sensitivity study conducted to identify the optimum reactor operating temperatures for producing the heat and hydrogen required for industrial processes associated with the proposed new high temperature gas-cooled reactor. This study assumed that primary steam outputs of the reactor were delivered at 17 MPa and 540°C and the helium coolant was delivered at 7 MPa at 625–925°C. The secondary outputs of were electricity and hydrogen. For the power generation analysis, it was assumed that the power cycle efficiency was 66% of the maximum theoretical efficiency of the Carnot thermodynamic cycle. Hydrogen was generated via the hightemperature steam electrolysis or the steam methane reforming process. The study indicates that optimum or a range of reactor outlet temperatures could be identified to further refine the process evaluations that were developed for high temperature gas-cooled reactor-integrated production of synthetic transportation fuels, ammonia, and ammonia derivatives, oil from unconventional sources, and substitute natural gas from coal.

  19. Hybrid adsorptive membrane reactor

    DOE Patents [OSTI]

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  20. H Reactor - Hanford Site

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    About Us Projects & Facilities H Reactor About Us About Hanford Cleanup Hanford History ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  1. C Reactor - Hanford Site

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    C Reactor About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  2. F Reactor - Hanford Site

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    About Us Projects & Facilities F Reactor About Us About Hanford Cleanup Hanford History ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  3. N Reactor - Hanford Site

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Projects & Facilities N Reactor About Us About Hanford Cleanup Hanford History Hanford ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  4. Control Means for Reactor

    DOE Patents [OSTI]

    Manley, J. H.

    1961-06-27

    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.

  5. Power Burst Facility (PBF) Reactor Reactor Decommissioning

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    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 (top), then places it ...

  6. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  7. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wigner, E.P.

    1957-09-17

    A reactor of the type having coolant liquid circulated through clad fuel elements geometrically arranged in a solid moderator, such as graphite, is described. The core is enclosed in a pressure vessel and suitable shielding, wherein means is provided for circulating vapor through the core to superheat the same. This is accomplished by drawing off the liquid which has been heated in the core due to the fission of the fuel, passing it to a nozzle within a chamber where it flashes into a vapor, and then passing the vapor through separate tubes extending through the moderator to pick up more heat developed in the core due to the fission of the fuel, thereby producing superheated vapor.

  8. Nuclear reactor

    DOE Patents [OSTI]

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  9. Oxidation of PCEA nuclear graphite by low water concentrations in helium

    SciTech Connect (OSTI)

    Contescu, Cristian I; Mee, Robert; Wang, Peng; Romanova, Anna V; Burchell, Timothy D

    2014-10-01

    Accelerated oxidation tests were performed to determine kinetic parameters of the chronic oxidation reaction of PCEA graphite in contact with helium coolant containing low moisture concentrations in high temperature gas-cooled reactors. To the authors best knowledge such a study has not been done since the detailed analysis of reaction of H-451 graphite with steam [Velasquez, Hightower, Burnette, 1978]. Since that H-451 graphite is now unavailable, it is urgently needed to characterize chronic oxidation behavior of new graphite grades under qualification for gas-cooled reactors. The Langmuir-Hinshelwood mechanism of carbon oxidation by water results in a non-linear reaction rate expression, with at least six different parameters. They were determined in accelerated oxidation experiments that covered a large range of temperatures (800 to 1100 oC), and partial pressures of water (15 to 850 Pa) and hydrogen (30 to 150 Pa) and used graphite specimens thin enough (4 mm) in order to avoid diffusion effects. Data analysis employed a statistical method based on multiple likelihood estimation of parameters and simultaneous fitting of non-linear equations. The results show significant material-specific differences between graphite grades PCEA and H-451 which were attributed to microstructural dissimilarity of the two materials. It is concluded that kinetic data cannot be transferred from one graphite grade to another.

  10. Process and installation for purification of the helium contained in a mixture of gas

    SciTech Connect (OSTI)

    Avon, M.F.; Markarian, G.R.

    1984-04-24

    The present invention relates to a process and an installation for purification of the helium contained in a mixture of gas, employing a pre-treatment unit to retain the impurities such as water, carbon dioxide gas and heavy organic compounds, and at least one reactor of the chromatographic type located downstream of said pre-treatment unit, said process comprising the following steps of: (a) adjusting the pressure of the mixture of gas until the working pressure of the phase of adsorption is obtained, this pressure being between 10 and 30 bars, and preferably 12 to 15 bars; (b) taking the temperature of the mixture of gas at the outlet of said pre-treatment unit until it is located in the range -15/sup 0/ C./-35/sup 0/ C., and preferably -25/sup 0/ C.; (c) and sending the mixture of gas into the reactor and passing it through an absorbent, which is constituted by a microporous charcoal whose pores are of dimensions less than or equal to 20 A.

  11. Medium-size high-temperature gas-cooled reactor

    SciTech Connect (OSTI)

    Peinado, C.O.; Koutz, S.L.

    1980-08-01

    This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760/sup 0/C (1400/sup 0/F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics (a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant) and engineered safety features (core auxiliary cooling, relief valve, and steam generator dump systems).

  12. An effective loading method of americium targets in fast reactors

    SciTech Connect (OSTI)

    Ohki, Shigeo; Sato, Isamu; Mizuno, Tomoyasu; Hayashi, Hideyuki; Tanaka, Kenya

    2007-07-01

    Recently, the development of target fuel with high americium (Am) content has been launched for the reduction of the overall fuel fabrication cost of the minor actinide (MA) recycling. In the framework of the development, this study proposes an effective loading method of Am targets in fast reactors. As a result of parametric survey calculations, we have found the ring-shaped target loading pattern between inner and outer core regions. This loading method is satisfactory both in core characteristics and in MA transmutation property. It should be noted that the Am targets can contribute to the suppression of the core power distribution change due to burnup. The major drawback of Am target is the production of helium gas. A target design modification by increasing the cladding thickness is found to be the most feasible measure to cope with the helium production. (authors)

  13. Helium gas bubble trapped in liquid helium in high magnetic field

    SciTech Connect (OSTI)

    Bai, H. Hannahs, S. T.; Markiewicz, W. D.; Weijers, H. W.

    2014-03-31

    High magnetic field magnets are used widely in the area of the condensed matter physics, material science, chemistry, geochemistry, and biology at the National High Magnetic Field Laboratory. New high field magnets of state-of-the-art are being pursued and developed at the lab, such as the current developing 32 T, 32 mm bore fully superconducting magnet. Liquid Helium (LHe) is used as the coolant for superconducting magnets or samples tested in a high magnetic field. When the magnetic field reaches a relatively high value the boil-off helium gas bubble generated by heat losses in the cryostat can be trapped in the LHe bath in the region where BzdBz/dz is less than negative 2100 T{sup 2}/m, instead of floating up to the top of LHe. Then the magnet or sample in the trapped bubble region may lose efficient cooling. In the development of the 32 T magnet, a prototype Yttrium Barium Copper Oxide coil of 6 double pancakes with an inner diameter of 40 mm and an outer diameter of 140 mm was fabricated and tested in a resistive magnet providing a background field of 15 T. The trapped gas bubble was observed in the tests when the prototype coil was ramped up to 7.5 T at a current of 200 A. This letter reports the test results on the trapped gas bubble and the comparison with the analytical results which shows they are in a good agreement.

  14. Reactor and method of operation

    DOE Patents [OSTI]

    Wheeler, John A.

    1976-08-10

    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.

  15. Controlled Chemistry Helium High Temperature Materials Test Loop

    SciTech Connect (OSTI)

    Richard N. WRight

    2005-08-01

    A system to test aging and environmental effects in flowing helium with impurity content representative of the Next Generation Nuclear Plant (NGNP) has been designed and assembled. The system will be used to expose microstructure analysis coupons and mechanical test specimens for up to 5,000 hours in helium containing potentially oxidizing or carburizing impurities controlled to parts per million levels. Impurity levels in the flowing helium are controlled through a feedback mechanism based on gas chromatography measurements of the gas chemistry at the inlet and exit from a high temperature retort containing the test materials. Initial testing will focus on determining the nature and extent of combined aging and environmental effects on microstructure and elevated temperature mechanical properties of alloys proposed for structural applications in the NGNP, including Inconel 617 and Haynes 230.

  16. Multi-Purpose Thermal Hydraulic Loop: Advanced Reactor Technology Integral System Test (ARTIST) Facility for Support of Advanced Reactor Technologies

    SciTech Connect (OSTI)

    James E. O'Brien; Piyush Sabharwall; SuJong Yoon

    2001-11-01

    Effective and robust high temperature heat transfer systems are fundamental to the successful deployment of advanced reactors for both power generation and non-electric applications. Plant designs often include an intermediate heat transfer loop (IHTL) with heat exchangers at either end to deliver thermal energy to the application while providing isolation of the primary reactor system. In order to address technical feasibility concerns and challenges a new high-temperature multi-fluid, multi-loop test facility “Advanced Reactor Technology Integral System Test facility” (ARTIST) is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water. Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed

  17. Light Water Reactor Sustainability (LWRS) Program | Department...

    Energy Savers

    Nuclear Reactor Technologies Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) ...

  18. Operation of drift chambers with helium based mixtures

    SciTech Connect (OSTI)

    Sharma, A. ); Sauli, F. )

    1994-08-01

    Helium based gas mixtures have been investigated for lowering multiple scattering contributions to the momentum resolution for intermediate energy particles. The relevant transport parameters, namely drift velocity and diffusion have been calculated for several mixtures and compared to standard argon based mixtures. Some fast, low diffusion mixtures have been identified. The small Lorentz angle computed make them promising candidates for drift chamber operation in magnetic fields. Measurements on high accuracy drift chambers in a test beam with a helium-DME (dimethyl ether) (70-30) mixture have resulted in a spatial resolution ranging from 70[mu]m to 100[mu]m.

  19. Isotope Effects and Helium Retention Behavior in Vanadium Tritide

    SciTech Connect (OSTI)

    Bowman, Jr., R. C.; Attalla, A.; Craft, B. D.

    1985-04-01

    The relaxation times of the H, T, and 3He nuclei have been measured in vanadium hydride and tritide samples. Substantial isotope effects in both the phase transition temperatures and diffusion parameters have been found. When compared to hydrides, the tritide samples have lower transition temperatures and faster mobilities. The differences in the occupancies of the interstitial sites are largely responsible for these isotope effects. Most of the helium atoms generated by tritium decay remain trapped in microscopic bubbles formed with the VTx lattice. Evidence is presented for the gradual growth of the helium bubbles over periods of hundreds of days.

  20. Status of the helium afterglow injector for MAMI

    SciTech Connect (OSTI)

    Arianer, J.; Cohen, S.; Essabaa, S.; Frascaria, R.; Zerhouni, O.; Szeremeta, F.; Wurzinger, R.; Aulenbacher, K.

    1998-01-20

    For the parity-violation experiment at the Mami cw electron accelerator facility at the University of Mainz, we need a stable polarized electron source able to deliver a current >150 {mu}A with polarization >60%. We develop the Orsay polarized electron source SELPO as an injector for MAMI. SELPO source based on a flowing helium afterglow is now operating in a high voltage platform. The terminal will be tested completely at Orsay and is foreseen to be installed at MAMI before the end of this year. Recent results obtained with an array of capillaries to generate the helium flow and a longitudinal spin direction are described.

  1. SRS Small Modular Reactors

    ScienceCinema (OSTI)

    None

    2014-05-21

    The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

  2. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J.

    1980-03-11

    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.

  3. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B.

    2004-03-16

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  4. Statistical properties of inter-series mixing in helium: From integrability to chaos

    SciTech Connect (OSTI)

    Puttner, R; Gremaud, B.; Delande, D.; Domke, M.; Martins, M.; Schlachter, A.S.; Kaindl, G.

    2001-04-23

    The photoionization spectrum of helium near the double-ionization threshold shows structure which indicated a transition towards quantum chaos.

  5. Laser induced fluorescence spectroscopy of the Ca dimer deposited on helium and mixed helium/xenon clusters

    SciTech Connect (OSTI)

    Gaveau, Marc-André; Pothier, Christophe; Briant, Marc; Mestdagh, Jean-Michel

    2014-12-09

    We study how the laser induced fluorescence spectroscopy of the calcium dimer deposited on pure helium clusters is modified by the addition of xenon atoms. In the wavelength range between 365 and 385 nm, the Ca dimer is excited from its ground state up to two excited electronic states leading to its photodissociation in Ca({sup 1}P)+Ca({sup 1}S): this process is monitored by recording the Ca({sup 1}P) fluorescence at 422.7nm. One of these electronic states of Ca{sub 2} is a diexcited one correlating to the Ca(4s4p{sup 3}P(+Ca(4s3d{sup 3}D), the other one is a repulsive state correlating to the Ca(4s4p1P)+Ca(4s21S) asymptote, accounting for the dissociation of Ca{sub 2} and the observation of the subsequent Ca({sup 1}P) emission. On pure helium clusters, the fluorescence exhibits the calcium atomic resonance line Ca({sup 1}S←{sup 1}P) at 422.7 nm (23652 cm{sup −1}) assigned to ejected calcium, and a narrow red sided band corresponding to calcium that remains solvated on the helium cluster. When adding xenon atoms to the helium clusters, the intensity of these two features decreases and a new spectral band appears on the red side of calcium resonance line; the intensity and the red shift of this component increase along with the xenon quantity deposited on the helium cluster: it is assigned to the emission of Ca({sup 1}P) associated with the small xenon aggregate embedded inside the helium cluster.

  6. THE INTEGRATION OF PROCESS HEAT APPLICATIONS TO HIGH TEMPERATURE GAS REACTORS

    SciTech Connect (OSTI)

    Michael G. McKellar

    2011-11-01

    A high temperature gas reactor, HTGR, can produce industrial process steam, high-temperature heat-transfer gases, and/or electricity. In conventional industrial processes, these products are generated by the combustion of fossil fuels such as coal and natural gas, resulting in significant emissions of greenhouse gases such as carbon dioxide. Heat or electricity produced in an HTGR could be used to supply process heat or electricity to conventional processes without generating any greenhouse gases. Process heat from a reactor needs to be transported by a gas to the industrial process. Two such gases were considered in this study: helium and steam. For this analysis, it was assumed that steam was delivered at 17 MPa and 540 C and helium was delivered at 7 MPa and at a variety of temperatures. The temperature of the gas returning from the industrial process and going to the HTGR must be within certain temperature ranges to maintain the correct reactor inlet temperature for a particular reactor outlet temperature. The returning gas may be below the reactor inlet temperature, ROT, but not above. The optimal return temperature produces the maximum process heat gas flow rate. For steam, the delivered pressure sets an optimal reactor outlet temperature based on the condensation temperature of the steam. ROTs greater than 769.7 C produce no additional advantage for the production of steam.

  7. An investigation of thermally driven acoustical oscillations in helium systems

    SciTech Connect (OSTI)

    Fuerst, J.D.

    1990-08-01

    The phenomenon of thermal-acoustic oscillation is seen to arise spontaneously in gas columns subjected to steep temperature gradients, particularly in tubes connecting liquid helium reservoirs with the ambient environment. This if often the arrangement for installed cryogenic instrumentation and is accompanied by undesirably large heat transfer rates to the cold region. Experimental data are collected and matched to theoretical predictions of oscillatory behavior; these results are in good agreement with the analytical model and with previously collected data. The present experiment places the open ends of oscillating tubes of the various lengths and cross sections in communication with flowing helium in the subcooled, 2-phase, or superheated state while the other ends are maintained at some controlled, elevated temperature. Assorted cold end conditions are achieved through adjustments to the Fermilab Tevatron satellite test refrigerator to which the test cryostat is connected. The warm, closed ends of the tubes are maintained by isothermal baths of liquid nitrogen, ice water, and boiling water. The method is contrasted to previous arrangements whereby tubes are run from room temperature into or adjacent to a stagnant pool of liquid helium. Additionally, the effect of pulsations in the flowing helium stream is explored through operation of the refrigerator's wet and dry expanders during data collection. These data confirm the theory to which try were compared and support its use in the design of cryogenic sensing lines for avoidance of thermoacoustic oscillation.

  8. TRANSMISSION ELECTRON MICROSCOPY STUDY OF HELIUM BEARING FUSION WELDS

    SciTech Connect (OSTI)

    Tosten, M; Michael Morgan, M

    2008-12-12

    A transmission electron microscopy (TEM) study was conducted to characterize the helium bubble distributions in tritium-charged-and-aged 304L and 21Cr-6Ni-9Mn stainless steel fusion welds containing approximately 150 appm helium-3. TEM foils were prepared from C-shaped fracture toughness test specimens containing {delta} ferrite levels ranging from 4 to 33 volume percent. The weld microstructures in the low ferrite welds consisted mostly of austenite and discontinuous, skeletal {delta} ferrite. In welds with higher levels of {delta} ferrite, the ferrite was more continuous and, in some areas of the 33 volume percent sample, was the matrix/majority phase. The helium bubble microstructures observed were similar in all samples. Bubbles were found in the austenite but not in the {delta} ferrite. In the austenite, bubbles had nucleated homogeneously in the grain interiors and heterogeneously on dislocations. Bubbles were not found on any austenite/austenite grain boundaries or at the austenite/{delta} ferrite interphase interfaces. Bubbles were not observed in the {delta} ferrite because of the combined effects of the low solubility and rapid diffusion of tritium through the {delta} ferrite which limited the amount of helium present to form visible bubbles.

  9. A review of existing gas-cooled reactor circulators with application of the lessons learned to the new production reactor circulators

    SciTech Connect (OSTI)

    White, L.S.

    1990-07-01

    This report presents the results of a study of the lessons learned during the design, testing, and operation of gas-cooled reactor coolant circulators. The intent of this study is to identify failure modes and problem areas of the existing circulators so this information can be incorporated into the design of the circulators for the New Production Reactor (NPR)-Modular High-Temperature Gas Cooled Reactor (MHTGR). The information for this study was obtained primarily from open literature and includes data on high-pressure, high-temperature helium test loop circulators as well as the existing gas cooled reactors worldwide. This investigation indicates that trouble free circulator performance can only be expected when the design program includes a comprehensive prototypical test program, with the results of this test program factored into the final circulator design. 43 refs., 7 tabs.

  10. Effect on the tritium breeding ratio for a distributed ICRF antenna in a DEMO reactor

    SciTech Connect (OSTI)

    Garcia, A.; Noterdaeme, J.-M.; Fischer, U.; Dies, J.

    2015-12-10

    The paper reports results of MCNP-5 calculations to assess the effect on the Tritium Breeding Ratio (TBR) when integrating a distributed Ion Cyclotron Range of Frequencies (ICRF) antenna in the blanket of DEMO fusion power reactor. The calculations consider different parameters such as the ICRF covering ratio and the type of breeding blanket including the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL) concepts. For an antenna with a full toroidal circumference of 360°, located poloidally at 40° with a poloidal extension of 1 m, the reduction of the TBR is −0.349% for the HCPB blanket and −0.532% for the HCLL blanket. The distributed ICRF antenna is thus shown to have only a marginal effect on the TBR of the DEMO reactor.

  11. Characterization of the products and comparison of the product yields from the flash pyrolysis of fir wood in hydrogen and helium

    SciTech Connect (OSTI)

    Sundaram, M.S.; Steinberg, M.; Fallon, P.T.

    1984-01-01

    A seasoned sawdust of Douglas Fir wood was flash pyrolyzed in an entrained tubular reactor in the presence of hydrogen and helium at short residence times (less than 4 sec) at temperatures varying from 600/sup 0/ to 1000/sup 0/C and pressures of 50 to 500 psi. A significant quantity of gaseous and liquid hydrocarbons and CO were produced. The liquid products were characterized via GC/MS and Pyrolysis Mass Spectrometry. The composition of the liquid products and the influence of the processing conditions on the product yields are discussed. 2 references, 13 figures, 2 tables.

  12. HTGR (High Temperature Gas-Cooled Reactor) ingress analysis using MINET

    SciTech Connect (OSTI)

    Van Tuyle, G.J.; Yang, J.W.; Kroeger, P.G.; Mallen, A.N.; Aronson, A.L.

    1989-04-01

    Modeling of water/steam ingress into the primary (helium) cooling circuit of a High Temperature Gas-Cooled Reactor (HTGR) is described. This modeling was implemented in the MINET Code, which is a program for analyzing transients in intricate fluid flow and heat transfer networks. Results from the simulation of a water ingress event postulated for the Modular HTGR are discussed. 27 refs., 6 figs., 6 tabs.

  13. Attrition reactor system

    DOE Patents [OSTI]

    Scott, C.D.; Davison, B.H.

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  14. Attrition reactor system

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  15. Hybrid plasmachemical reactor

    SciTech Connect (OSTI)

    Lelevkin, V. M. Smirnova, Yu. G.; Tokarev, A. V.

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  16. Reactor System Transient Code.

    Energy Science and Technology Software Center (OSTI)

    1999-07-14

    RELAP3B describes the behavior of water-cooled nuclear reactors during postulated accidents or power transients, such as large reactivity excursions, coolant losses or pump failures. The program calculates flows, mass and energy inventories, pressures, temperatures, and steam qualities along with variables associated with reactor power, reactor heat transfer, or control systems. Its versatility allows one to describe simple hydraulic systems as well as complex reactor systems.

  17. Period meter for reactors

    DOE Patents [OSTI]

    Rusch, Gordon K.

    1976-01-06

    An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

  18. NEUTRONIC REACTOR POWER PLANT

    DOE Patents [OSTI]

    Metcalf, H.E.

    1962-12-25

    This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

  19. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    SciTech Connect (OSTI)

    Shropshire, D.E.; Herring, J.S.

    2004-10-03

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

  20. NEUTRONIC REACTOR SHIELDING

    DOE Patents [OSTI]

    Borst, L.B.

    1961-07-11

    A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

  1. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28

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

  2. High solids fermentation reactor

    DOE Patents [OSTI]

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

    1993-01-01

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

  3. High solids fermentation reactor

    DOE Patents [OSTI]

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

    1993-03-02

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

  4. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, James P.; Scahill, John W.

    1995-01-01

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  5. Advanced Test Reactor Tour

    SciTech Connect (OSTI)

    Miley, Don

    2011-01-01

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

  6. NUCLEAR REACTOR CONTROL SYSTEM

    DOE Patents [OSTI]

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

  7. Source localization of brain activity using helium-free interferometer

    SciTech Connect (OSTI)

    Dammers, Jürgen Chocholacs, Harald; Eich, Eberhard; Boers, Frank; Faley, Michael; Dunin-Borkowski, Rafal E.; Jon Shah, N.

    2014-05-26

    To detect extremely small magnetic fields generated by the human brain, currently all commercial magnetoencephalography (MEG) systems are equipped with low-temperature (low-T{sub c}) superconducting quantum interference device (SQUID) sensors that use liquid helium for cooling. The limited and increasingly expensive supply of helium, which has seen dramatic price increases recently, has become a real problem for such systems and the situation shows no signs of abating. MEG research in the long run is now endangered. In this study, we report a MEG source localization utilizing a single, highly sensitive SQUID cooled with liquid nitrogen only. Our findings confirm that localization of neuromagnetic activity is indeed possible using high-T{sub c} SQUIDs. We believe that our findings secure the future of this exquisitely sensitive technique and have major implications for brain research and the developments of cost-effective multi-channel, high-T{sub c} SQUID-based MEG systems.

  8. Quantum entanglement for helium atom in the Debye plasmas

    SciTech Connect (OSTI)

    Lin, Yen-Chang; Fang, Te-Kuei; Ho, Yew Kam

    2015-03-15

    In the present work, we present an investigation on quantum entanglement of the two-electron helium atom immersed in weakly coupled Debye plasmas, modeled by the Debye-Hckel, or screened Coulomb, potential to mimic the interaction between two charged particles inside the plasma. Quantum entanglement is related to correlation effects in a multi-particle system. In a bipartite system, a measurement made on one of the two entangled particles affects the outcome of the other particle, even if such two particles are far apart. Employing wave functions constructed with configuration interaction B-spline basis, we have quantified von Neumann entropy and linear entropy for a series of He {sup 1,3}S{sup e} and {sup 1,3}P{sup o} states in plasma-embedded helium atom.

  9. Effect of Helium Accumulation on the Spent Fuel Microstructure

    SciTech Connect (OSTI)

    Ferry, Cecile; Piron, Jean-Paul; Stout, Ray

    2007-07-01

    In a nuclear spent fuel repository, the aqueous rapid release of radio-activity from exposed spent fuel surfaces will depend on the pellet microstructure at the arrival time of water into the disposal container. Research performed on spent fuel evolution in a closed system has shown that the evolution of microstructure under disposal conditions should be governed by the cumulated {alpha}-decay damage and the subsequent helium behavior. The evolution of fission gas bubble characteristics under repository conditions has to be assessed. In UO{sub 2} fuels with a burnup of 47.5 GWd/t, the pressure in fission gas bubbles, including the pressure increase from {alpha}-decay helium atoms, is not expected to reach the critical bubble pressure that will cause failure, thus micro-cracking in UO{sub 2} spent fuel grains is not expected. (authors)

  10. Elastic Electron Scattering from Tritium and Helium-3

    DOE R&D Accomplishments [OSTI]

    Collard, H.; Hofstadter, R.; Hughes, E. B.; Johansson, A.; Yearian, M. R.; Day, R. B.; Wagner, R. T.

    1964-10-01

    The mirror nuclei of tritium and helium-3 have been studied by the method of elastic electron scattering. Absolute cross sections have been measured for incident electron energies in the range 110 - 690 MeV at scattering angles lying between 40 degrees and 135 degrees in this energy range. The data have been interpreted in a straightforward manner and form factors are given for the distributions of charge and magnetic moment in the two nuclei over a range of four-momentum transfer squared 1.0 - 8.0 F{sup -2}. Model-independent radii of the charge and magnetic moment distributions are given and an attempt is made to deduce form factors describing the spatial distribution of the protons in tritium and helium-3.

  11. Investigation of Cellular Interactions of Nanoparticles by Helium Ion Microscopy

    SciTech Connect (OSTI)

    Arey, Bruce W.; Shutthanandan, V.; Xie, Yumei; Tolic, Ana; Williams, Nolann G.; Orr, Galya

    2011-06-01

    The helium ion mircroscope (HIM) probes light elements (e.g. C, N, O, P) with high contrast due to the large variation in secondary electron yield, which minimizes the necessity of specimen staining. A defining characteristic of HIM is its remarkable capability to neutralize charge by the implementation of an electron flood gun, which eliminates the need for coating non-conductive specimens for imaging at high resolution. In addition, the small convergence angle in HeIM offers a large depth of field (~5x FE-SEM), enabling tall structures to be viewed in focus within a single image. Taking advantage of these capabilities, we investigate the interactions of engineered nanoparticles (NPs) at the surface of alveolar type II epithelial cells grown at the air-liquid interface (ALI). The increasing use of nanomaterials in a wide range of commercial applications has the potential to increase human exposure to these materials, but the impact of such exposure on human health is still unclear. One of the main routs of exposure is the respiratory tract, where alveolar epithelial cells present a vulnerable target at the interface with ambient air. Since the cellular interactions of NPs govern the cellular response and ultimately determine the impact on human health, our studies will help delineating relationships between particle properties and cellular interactions and response to better evaluate NP toxicity or biocompatibility. The Rutherford backscattered ion (RBI) is a helium ions imaging mode, which backscatters helium ions from every element except hydrogen, with a backscatter yield that depends on the atomic number of the target. Energy-sensitive backscatter analysis is being developed, which when combined with RBI image information, supports elemental identification at helium ion nanometer resolution. This capability will enable distinguishing NPs from cell surface structures with nanometer resolution.

  12. Mathematical modeling of a Fermilab helium liquefier coldbox

    SciTech Connect (OSTI)

    Geynisman, M.G.; Walker, R.J.

    1995-12-01

    Fermilab Central Helium Liquefier (CHL) facility is operated 24 hours-a-day to supply 4.6{degrees}K for the Fermilab Tevatron superconducting proton-antiproton collider Ring and to recover warm return gases. The centerpieces of the CHL are two independent cold boxes rated at 4000 and 5400 liters/hour with LN{sub 2} precool. These coldboxes are Claude cycle and have identical heat exchangers trains, but different turbo-expanders. The Tevatron cryogenics demand for higher helium supply from CHL was the driving force to investigate an installation of an expansion engine in place of the Joule-Thompson valve. A mathematical model was developed to describe the thermo- and gas-dynamic processes for the equipment included in the helium coldbox. The model is based on a finite element approach, opposite to a global variables approach, thus providing for higher accuracy and conversion stability. Though the coefficients used in thermo- and gas-dynamic equations are unique for a given coldbox, the general approach, the equations, the methods of computations, and most of the subroutines written in FORTRAN can be readily applied to different coldboxes. The simulation results are compared against actual operating data to demonstrate applicability of the model.

  13. Reactor vessel support system

    DOE Patents [OSTI]

    Golden, Martin P.; Holley, John C.

    1982-01-01

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  14. Nuclear reactor overflow line

    DOE Patents [OSTI]

    Severson, Wayne J.

    1976-01-01

    The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

  15. B Reactor - Hanford Site

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Reactor About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs Contact Us 100 Area 118-K-1 Burial Ground 200 Area 222-S Laboratory 242-A Evaporator 300 Area 324 Building 325 Building 400 Area/Fast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim Storage Area Canyon Facilities Cold Test Facility D and DR Reactors Effluent Treatment Facility Environmental Restoration Disposal Facility F Reactor H

  16. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    SciTech Connect (OSTI)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-05-01

    The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers.

  17. Spinning fluids reactor

    DOE Patents [OSTI]

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  18. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  19. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  20. Fission reactors and materials

    SciTech Connect (OSTI)

    Frost, B.R.T.

    1981-12-01

    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.

  1. Helium-Based Soundwave Chiller: Trillium: A Helium-Based Sonic Chiller- Tons of Freezing with 0 GWP Refrigerants

    SciTech Connect (OSTI)

    2010-09-01

    BEETIT Project: Penn State is designing a freezer that substitutes the use of sound waves and environmentally benign refrigerant for synthetic refrigerants found in conventional freezers. Called a thermoacoustic chiller, the technology is based on the fact that the pressure oscillations in a sound wave result in temperature changes. Areas of higher pressure raise temperatures and areas of low pressure decrease temperatures. By carefully arranging a series of heat exchangers in a sound field, the chiller is able to isolate the hot and cold regions of the sound waves. Penn State’s chiller uses helium gas to replace synthetic refrigerants. Because helium does not burn, explode or combine with other chemicals, it is an environmentally-friendly alternative to other polluting refrigerants. Penn State is working to apply this technology on a large scale.

  2. Pulsed deuterium lithium nuclear reactor

    SciTech Connect (OSTI)

    Fischer, A.G.

    1980-01-08

    A nuclear reactor that burns hydrogen bomb material 6-lithium deuterotritide to helium in successive microexplosions which are ignited electrically and enclosed by this same molten material, and that permits the conversion of the reaction heat into useful electrical power. A specially-constructed high-current pulse machine is discharged via a thermally-preformed highly conducting path through a mass of the molten salt 6lid1-xtx (0

  3. Asteroseismic estimate of helium abundance of a solar analog binary system

    SciTech Connect (OSTI)

    Verma, Kuldeep; Antia, H. M.; Faria, Joo P.; Monteiro, Mrio J. P. F. G.; Basu, Sarbani; Mazumdar, Anwesh; Appourchaux, Thierry; Chaplin, William J.; Garca, Rafael A.

    2014-08-01

    16 Cyg A and B are among the brightest stars observed by Kepler. What makes these stars more interesting is that they are solar analogs. 16 Cyg A and B exhibit solar-like oscillations. In this work we use oscillation frequencies obtained using 2.5 yr of Kepler data to determine the current helium abundance of these stars. For this we use the fact that the helium ionization zone leaves a signature on the oscillation frequencies and that this signature can be calibrated to determine the helium abundance of that layer. By calibrating the signature of the helium ionization zone against models of known helium abundance, the helium abundance in the envelope of 16 Cyg A is found to lie in the range of 0.231 to 0.251 and that of 16 Cyg B lies in the range of 0.218 to 0.266.

  4. Production of carbon monoxide-free hydrogen and helium from a high-purity source

    DOE Patents [OSTI]

    Golden, Timothy Christopher; Farris, Thomas Stephen

    2008-11-18

    The invention provides vacuum swing adsorption processes that produce an essentially carbon monoxide-free hydrogen or helium gas stream from, respectively, a high-purity (e.g., pipeline grade) hydrogen or helium gas stream using one or two adsorber beds. By using physical adsorbents with high heats of nitrogen adsorption, intermediate heats of carbon monoxide adsorption, and low heats of hydrogen and helium adsorption, and by using vacuum purging and high feed stream pressures (e.g., pressures of as high as around 1,000 bar), pipeline grade hydrogen or helium can purified to produce essentially carbon monoxide -free hydrogen and helium, or carbon monoxide, nitrogen, and methane-free hydrogen and helium.

  5. Interim Report on the Optimization and Feasibility Studies for the Neutron Detection without Helium-3 Project

    SciTech Connect (OSTI)

    Ely, James H.

    2012-03-19

    This report provides the status and results of the first year's effort in modeling and simulation to investigate alternatives to helium-3 for neutron detection in safeguards applications.

  6. Ultra high vacuum pumping system and high sensitivity helium leak detector

    DOE Patents [OSTI]

    Myneni, G.R.

    1997-12-30

    An improved helium leak detection method and apparatus are disclosed which increase the leak detection sensitivity to 10{sup {minus}13} atm cc/s. The leak detection sensitivity is improved over conventional leak detectors by completely eliminating the use of o-rings, equipping the system with oil-free pumping systems, and by introducing measured flows of nitrogen at the entrances of both the turbo pump and backing pump to keep the system free of helium background. The addition of dry nitrogen flows to the system reduces back streaming of atmospheric helium through the pumping system as a result of the limited compression ratios of the pumps for helium. 2 figs.

  7. Energy-related applications of helium: a revision of the ERDA-13 data base

    SciTech Connect (OSTI)

    Hammel, E.F.; Krupka, M.C.

    1980-08-01

    A re-examination, revision, and re-evaluation of the data base contained within the 1975 document, ERDA-13, The Energy-Related Applications of Helium, were completed and results are presented in this report. New technical and resource data, current legislative proposals, updated supply-and-demand relationships, latest legal developments, programmatic changes affectng the future demand for helium, socio-economic aspects, and the effects of the latest energy-consumption projections were considered and are discussed. In contrast to ERDA-13, however, explicit recommendations with respect to the formulation of Federal helium policy, as it pertains to the energy-related applications of helium, are not given.

  8. Ultra high vacuum pumping system and high sensitivity helium leak detector

    DOE Patents [OSTI]

    Myneni, Ganapati Rao

    1997-01-01

    An improved helium leak detection method and apparatus are disclosed which increase the leak detection sensitivity to 10.sup.-13 atm cc s.sup.-1. The leak detection sensitivity is improved over conventional leak detectors by completely eliminating the use of o-rings, equipping the system with oil-free pumping systems, and by introducing measured flows of nitrogen at the entrances of both the turbo pump and backing pump to keep the system free of helium background. The addition of dry nitrogen flows to the system reduces backstreaming of atmospheric helium through the pumping system as a result of the limited compression ratios of the pumps for helium.

  9. The Other Helium Shortage Smart Telescopes Go Stargazing No-bang...

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    ... Less well known, however, is a more severe shortage of the much rarer isotope helium-3. Essential for key scientific, medical, and national security applications, no one would ...

  10. Theory of Positron Annihilation in Helium-Filled Bubbles in Plutonium...

    Office of Scientific and Technical Information (OSTI)

    Theory of Positron Annihilation in Helium-Filled Bubbles in Plutonium Citation Details ... This method is capable of treating system cell sizes of several thousand atoms, allowing ...

  11. NEUTRONIC REACTOR SYSTEM

    DOE Patents [OSTI]

    Goett, J.J.

    1961-01-24

    A system is described which includes a neutronic reactor containing a dispersion of fissionable material in a liquid moderator as fuel and a conveyor to which a portion of the dispersion may be passed and wherein the self heat of the slurry evaporates the moderator. Means are provided for condensing the liquid moderator and returning it to the reactor and for conveying the dried fissionable material away from the reactor.

  12. THERMAL NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Spinrad, B.I.

    1960-01-12

    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.

  13. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, J.P.; Scahill, J.W.

    1995-05-09

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  14. Small Modular Reactors (SMRs) | Department of Energy

    Energy Savers

    Reactor Technologies Small Modular Reactors (SMRs) Small Modular Reactors (SMRs) ... to the NRC by late-2016 Complete reactor module final design by mid-2019 For more ...

  15. Baseline Concept Description of a Small Modular High Temperature Reactor

    SciTech Connect (OSTI)

    Hans Gougar

    2014-05-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  16. Baseline Concept Description of a Small Modular High Temperature Reactor

    SciTech Connect (OSTI)

    Gougar, Hans D.

    2014-10-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  17. Baseline Concept Description of a Small Modular High Temperature Reactor

    SciTech Connect (OSTI)

    Hans Gougar

    2014-05-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both small or medium-sized and modular by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOEs ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV

  18. Tokamak reactor first wall

    DOE Patents [OSTI]

    Creedon, R.L.; Levine, H.E.; Wong, C.; Battaglia, J.

    1984-11-20

    This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

  19. Small Modular Reactors - SRSCRO

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    smr Small Modular Reactors The Savannah River National Laboratory (SRNL) has announced several partnerships to bring refrigerator-sized modular nuclear reactors, known as Small Modular Reactors or SMRs, to the Savannah River Site facility and jump start development of the U.S. Energy Freedom CenterTM. Currently, all large commercial power reactors in the United States and most in the rest of the world are based on "light water" designs - that is, they use uranium fuel and ordinary

  20. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, Juhani

    1996-01-01

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

  1. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, J.

    1996-03-19

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

  2. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOE Patents [OSTI]

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  3. Helium measurements of pore-fluids obtained from SAFOD drillcore

    SciTech Connect (OSTI)

    Ali, S.; Stute, M.; Torgersen, T.; Winckler, G.; Kennedy, B.M.

    2010-04-15

    {sup 4}He accumulated in fluids is a well established geochemical tracer used to study crustal fluid dynamics. Direct fluid samples are not always collectable; therefore, a method to extract rare gases from matrix fluids of whole rocks by diffusion has been adapted. Helium was measured on matrix fluids extracted from sandstones and mudstones recovered during the San Andreas Fault Observatory at Depth (SAFOD) drilling in California, USA. Samples were typically collected as subcores or from drillcore fragments. Helium concentration and isotope ratios were measured 4-6 times on each sample, and indicate a bulk {sup 4}He diffusion coefficient of 3.5 {+-} 1.3 x 10{sup -8} cm{sup 2}s{sup -1} at 21 C, compared to previously published diffusion coefficients of 1.2 x 10{sup -18} cm{sup 2}s{sup -1} (21 C) to 3.0 x 10{sup -15} cm{sup 2}s{sup -1} (150 C) in the sands and clays. Correcting the diffusion coefficient of {sup 4}He{sub water} for matrix porosity ({approx}3%) and tortuosity ({approx}6-13) produces effective diffusion coefficients of 1 x 10{sup -8} cm{sup 2}s{sup -1} (21 C) and 1 x 10{sup -7} (120 C), effectively isolating pore fluid {sup 4}He from the {sup 4}He contained in the rock matrix. Model calculations indicate that <6% of helium initially dissolved in pore fluids was lost during the sampling process. Complete and quantitative extraction of the pore fluids provide minimum in situ porosity values for sandstones 2.8 {+-} 0.4% (SD, n=4) and mudstones 3.1 {+-} 0.8% (SD, n=4).

  4. Active helium target: Neutron scalar polarizability extraction via Compton scattering

    SciTech Connect (OSTI)

    Morris, Meg Hornidge, David; Annand, John; Strandberg, Bruno

    2015-12-31

    Precise measurement of the neutron scalar polarizabilities has been a lasting challenge because of the lack of a free-neutron target. Led by the University of Glasgow and the Mount Allison University groups of the A2 collaboration in Mainz, Germany, preparations have begun to test a recent theoretical model with an active helium target with the hope of determining these elusive quantities with small statistical, systematic, and model-dependent errors. Apparatus testing and background-event simulations have been carried out, with the full experiment projected to run in 2015. Once determined, these values can be applied to help understand quantum chromodynamics in the nonperturbative region.

  5. Dual-frequency glow discharges in atmospheric helium

    SciTech Connect (OSTI)

    Huang, Xiaojiang; Guo, Ying; Dai, Lu; Zhang, Jing; Shi, J. J.

    2015-10-15

    In this paper, the dual-frequency (DF) glow discharges in atmospheric helium were experimented by electrical and optical measurements in terms of current voltage characteristics and optical emission intensity. It is shown that the waveforms of applied voltages or discharge currents are the results of low frequency (LF) waveforms added to high frequency (HF) waveforms. The HF mainly influences discharge currents, and the LF mainly influences applied voltages. The gas temperatures of DF discharges are mainly affected by HF power rather than LF power.

  6. Helium nano-bubble evolution in aging metal tritides.

    SciTech Connect (OSTI)

    Cowgill, Donald F.

    2004-05-01

    A continuum-scale, evolutionary model of helium (He) nano-bubble nucleation, growth and He release for aging bulk metal tritides is presented which accounts for major features of the experimental database. Bubble nucleation, modeled as self-trapping of interstitially diffusing He atoms, is found to occur during the first few days following tritium introduction into the metal and is sensitive to the He diffusivity and pairing energy. An effective helium diffusivity of 0.3 x 10{sup -16} cm{sup 2}/s at 300 K is required to generate the average bubble density of 5x 1017 bubbles/cm3 observed by transmission electron microscopy (TEM). Early bubble growth by dislocation loop punching with a l/radius bubble pressure dependence produces good agreement with He atomic volumes and bubble pressures determined from swelling data, nuclear magnetic resonance (NMR) measurements, and hydride pressure-composition-temperature (PCT) shifts. The model predicts that later in life neighboring bubble interactions may first lower the loop punching pressure through cooperative stress effects, then raise the pressure by partial blocking of loops. It also accounts for the shape of the bubble spacing distribution obtained from NMR data. This distribution is found to remain fixed with age, justifying the separation of nucleation and growth phases, providing a sensitive test of the growth formulation, and indicating that further significant bubble nucleation does not occur throughout life. Helium generated within the escape depth of surfaces and surface-connected porosity produces the low-level early helium release. Accelerated or rapid release is modeled as inter-bubble fracture using an average ligament stress criterion. Good agreement is found between the predicted onset of fracture and the observed He-metal ratio (HeM) for rapid He release from bulk palladium tritide. An examination of how inter-bubble fracture varies over the bubble spacing distribution shows that the critical Hem will be

  7. Thermal hydraulic performance testing of printed circuit heat exchangers in a high-temperature helium test facility

    SciTech Connect (OSTI)

    Sai K. Mylavarapu; Xiaodong Sun; Richard E. Glosup; Richard N. Christensen; Michael W. Patterson

    2014-04-01

    In high-temperature gas-cooled reactors, such as a very high temperature reactor (VHTR), an intermediate heat exchanger (IHX) is required to efficiently transfer the core thermal output to a secondary fluid for electricity generation with an indirect power cycle and/or process heat applications. Currently, there is no proven high-temperature (750800 C or higher) compact heat exchanger technology for high-temperature reactor design concepts. In this study, printed circuit heat exchanger (PCHE), a potential IHX concept for high-temperature applications, has been investigated for their heat transfer and pressure drop characteristics under high operating temperatures and pressures. Two PCHEs, each having 10 hot and 10 cold plates with 12 channels (semicircular cross-section) in each plate are fabricated using Alloy 617 plates and tested for their performance in a high-temperature helium test facility (HTHF). The PCHE inlet temperature and pressure were varied from 85 to 390 C/1.02.7 MPa for the cold side and 208790 C/1.02.7 MPa for the hot side, respectively, while the mass flow rate of helium was varied from 15 to 49 kg/h. This range of mass flow rates corresponds to PCHE channel Reynolds numbers of 950 to 4100 for the cold side and 900 to 3900 for the hot side (corresponding to the laminar and laminar-to-turbulent transition flow regimes). The obtained experimental data have been analyzed for the pressure drop and heat transfer characteristics of the heat transfer surface of the PCHEs and compared with the available models and correlations in the literature. In addition, a numerical treatment of hydrodynamically developing and hydrodynamically fully-developed laminar flow through a semicircular duct is presented. Relations developed for determining the hydrodynamic entrance length in a semicircular duct and the friction factor (or pressure drop) in the hydrodynamic entry length region for laminar flow through a semicircular duct are given. Various hydrodynamic

  8. Brookhaven Graphite Research Reactor Workshop

    Energy.gov [DOE]

    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...

  9. NEUTRONIC REACTOR BURIAL ASSEMBLY

    DOE Patents [OSTI]

    Treshow, M.

    1961-05-01

    A burial assembly is shown whereby an entire reactor core may be encased with lead shielding, withdrawn from the reactor site and buried. This is made possible by a five-piece interlocking arrangement that may be easily put together by remote control with no aligning of bolt holes or other such close adjustments being necessary.

  10. NEUTRONIC REACTOR SYSTEM

    DOE Patents [OSTI]

    Treshow, M.

    1959-02-10

    A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

  11. Light water reactor program

    SciTech Connect (OSTI)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  12. REFLECTOR FOR NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Fraas, A.P.

    1963-08-01

    A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

  13. Ignition and extinction phenomena in helium micro hollow cathode discharges

    SciTech Connect (OSTI)

    Kulsreshath, M. K.; Schwaederle, L.; Dufour, T.; Lefaucheux, P.; Dussart, R.; Overzet, L. J.

    2013-12-28

    Micro hollow cathode discharges (MHCD) were produced using 250??m thick dielectric layer of alumina sandwiched between two nickel electrodes of 8??m thickness. A through cavity at the center of the chip was formed by laser drilling technique. MHCD with a diameter of few hundreds of micrometers allowed us to generate direct current discharges in helium at up to atmospheric pressure. A slowly varying ramped voltage generator was used to study the ignition and the extinction periods of the microdischarges. The analysis was performed by using electrical characterisation of the V-I behaviour and the measurement of He*({sup 3}S{sub 1}) metastable atoms density by tunable diode laser spectroscopy. At the ignition of the microdischarges, 2??s long current peak as high as 24?mA was observed, sometimes followed by low amplitude damped oscillations. At helium pressure above 400?Torr, an oscillatory behaviour of the discharge current was observed just before the extinction of the microdischarges. The same type of instability in the extinction period at high pressure also appeared on the density of He*({sup 3}S{sub 1}) metastable atoms, but delayed by a few ?s relative to the current oscillations. Metastable atoms thus cannot be at the origin of the generation of the observed instabilities.

  14. Gettering of Hydrogen and Methane from a Helium Gas Mixture

    DOE PAGES-Beta [OSTI]

    Cardenas, Rosa E.; Stewart, Kenneth D.; Cowgill, Donald F.

    2014-10-21

    In our study, the authors developed an approach for accurately quantifying the helium content in a gas mixture also containing hydrogen and methane using commercially available getters. The authors performed a systematic study to examine how both H2 and CH4 can be removed simultaneously from the mixture using two SAES St 172® getters operating at different temperatures. The remaining He within the gas mixture can then be measured directly using a capacitance manometer. Moreover, the optimum combination involved operating one getter at 650°C to decompose the methane, and the second at 110°C to remove the hydrogen. Finally, this approach eliminatedmore » the need to reactivate the getters between measurements, thereby enabling multiple measurements to be made within a short time interval, with accuracy better than 1%. The authors anticipate that such an approach will be particularly useful for quantifying the He-3 in mixtures that include tritium, tritiated methane, and helium-3. The presence of tritiated methane, generated by tritium activity, often complicates such measurements.« less

  15. Gettering of Hydrogen and Methane from a Helium Gas Mixture

    SciTech Connect (OSTI)

    Cardenas, Rosa E.; Stewart, Kenneth D.; Cowgill, Donald F.

    2014-10-21

    In our study, the authors developed an approach for accurately quantifying the helium content in a gas mixture also containing hydrogen and methane using commercially available getters. The authors performed a systematic study to examine how both H2 and CH4 can be removed simultaneously from the mixture using two SAES St 172® getters operating at different temperatures. The remaining He within the gas mixture can then be measured directly using a capacitance manometer. Moreover, the optimum combination involved operating one getter at 650°C to decompose the methane, and the second at 110°C to remove the hydrogen. Finally, this approach eliminated the need to reactivate the getters between measurements, thereby enabling multiple measurements to be made within a short time interval, with accuracy better than 1%. The authors anticipate that such an approach will be particularly useful for quantifying the He-3 in mixtures that include tritium, tritiated methane, and helium-3. The presence of tritiated methane, generated by tritium activity, often complicates such measurements.

  16. Gettering of hydrogen and methane from a helium gas mixture

    SciTech Connect (OSTI)

    Crdenas, Rosa Elia; Stewart, Kenneth D.; Cowgill, Donald F.

    2014-11-01

    In this study, the authors developed an approach for accurately quantifying the helium content in a gas mixture also containing hydrogen and methane using commercially available getters. The authors performed a systematic study to examine how both H{sub 2} and CH{sub 4} can be removed simultaneously from the mixture using two SAES St 172{sup } getters operating at different temperatures. The remaining He within the gas mixture can then be measured directly using a capacitance manometer. The optimum combination involved operating one getter at 650?C to decompose the methane, and the second at 110?C to remove the hydrogen. This approach eliminated the need to reactivate the getters between measurements, thereby enabling multiple measurements to be made within a short time interval, with accuracy better than 1%. The authors anticipate that such an approach will be particularly useful for quantifying the He-3 in mixtures that include tritium, tritiated methane, and helium-3. The presence of tritiated methane, generated by tritium activity, often complicates such measurements.

  17. Helium Nano-Bubble Evolution in Aging Metal Tritides

    SciTech Connect (OSTI)

    Cowgill, Donald F.

    2005-07-15

    A continuum-scale, evolutionary model of bubble nucleation, growth and He release for aging metal tritides is described which accounts for major features of the tritide database. Bubble nucleation, modeled as self-trapping of interstitially diffusing He atoms, occurs during the first few days following tritium introduction into the metal. Bubble growth by dislocation loop punching yields good agreement between He atomic volumes and bubble pressures determined from bulk swelling and {sup 3}He NMR data. The bubble spacing distribution determined from NMR is shown to remain fixed with age, justifying the separation of nucleation and growth phases and providing a sensitive test of the growth formulation. Late in life, bubble interactions are proposed to produce cooperative stress effects, which lower the bubble pressure. Helium generated near surfaces and surface-connected porosity accounts for the low-level early helium release. Use of an average ligament stress criterion predicts an onset of inter-bubble fracture in good agreement with the He/Metal ratio observed for rapid He release. From the model, it is concluded that He retention can be controlled through control of bubble nucleation.

  18. Partial cross sections of helium satellites at medium photon energies

    SciTech Connect (OSTI)

    Wehlitz, R.; Sellin, I.A.; Hemmers, O.

    1997-04-01

    Still of current interest is the important role of single ionization with excitation compared to single ionization alone. The coupling between the electrons and the incoming photon is a single-particle operator. Thus, an excitation in addition to an ionization, leading to a so-called satellite line in a photoelectron spectrum, is entirely due to electron-electron interaction and probes the electron correlation in the ground and final state. Therefore the authors have undertaken the study of the intensity of helium satellites He{sup +}nl (n = 2 - 6) relative to the main photoline (n = 1) as a function of photon energy at photon energies well above threshold up to 900 eV. From these results they could calculate the partial cross-sections of the helium satellites. In order to test the consistency of their satellite-to-1s ratios with published double-to-single photoionization ratios, the authors calculated the double-to-single photoionization ratio from their measured ratios using the theoretical energy-distribution curves of Chang and Poe and Le Rouzo and Dal Cappello which proved to be valid for photon energies below 120 eV. These calculated double-to-single ionization ratios agree fairly well with recent ion measurements. In the lower photon energy range the authors ratios agree better with the ratios of Doerner et al. while for higher photon energies the agreement is better with the values of Levin et al.

  19. Mass spectrometric helium analysis of solid and gas samples from cold-fusion type experiments

    SciTech Connect (OSTI)

    Oliver, B.M.

    1995-12-01

    A custom mass spectrometer system, operating in static mode, has been used to measure helium in both solid and gas samples front cold-fusion type experiments. The mass spectrometer is a 2-in. Radius, 60{degrees}, permanent angle magnet instrument with a single electron-multiplier collecting. Depending on the absolute levels of helium expected, the analysis are conducted by isotope dilution or by measuring absolute collector values. Solid samples are vaporized to ensure complete helium release. Prior to analysis, the fraction of sample gas to be analyzed is exposed to a series of physical and chemical getters, including room temperature Zr-Al alloy (SAES type 101) and liquid-nitrogen cooled activated charcoal. This is done to remove active gases and hydrogen isotopes which could interfere with the helium determinations. Generally, the analysis protocol is to analyze an equal or greater number of {open_quotes}controls{close_quotes} along with the samples to accurately characterize system background and reproducibility. Absolute sensitivity for the system is approximately 1 x 10{sup 9} atoms. Absolute accuracy is 1% or better for helium levels > 10{sup 11} atoms. With few exceptions, helium analysis of solid samples front cold fusion type experiments have yielded no excess helium above usual system background. A few samples have shown helium levels in the low 10{sup 9} atom range, and some gas samples have shown {sup 4}He levels up to several hundred ppm.

  20. Some Aspects of Reactor Theory

    DOE R&D Accomplishments [OSTI]

    Weinberg, Alvin M.

    1952-10-10

    Some general remarks are made on reactor theory, particularly the asymptotic theory and multigroup methods. Unsolved reactor problems are also briefly discussed. (B.J.H.)

  1. Reactor Materials | Department of Energy

    Energy.gov (indexed) [DOE]

    reactor materials crosscut effort will enable the development of innovative and ... Research into specific degradation modes or material needs unique to a particular reactor ...

  2. Fundamental and applied studies of helium ingrowth and aging in plutonium

    SciTech Connect (OSTI)

    Stevens, M.F.; Zocco, T.; Albers, R.; Becker, J.D.; Walter, K.; Cort, B.; Paisley, D.; Nastasi, M.

    1998-12-31

    This is the final report of a two-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The purpose of this project was to develop new capabilities to assess the nucleation and growth of helium-associated defects in aged plutonium metal. This effort involved both fundamental and applied models to assist in predicting the transport and kinetics of helium in the metal lattice as well as ab initio calculations of the disposition of gallium in the fcc plutonium lattice and its resulting effects on phase stability. Experimentally this project aimed to establish experimental capabilities crucial to the prediction of helium effects in metals, such as transmission electron microscopy, thermal helium effusion, and the development of a laser-driven mini-flyer for understanding the role of helium and associated defects on shock response of plutonium surrogates.

  3. DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

    SciTech Connect (OSTI)

    Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

    2003-11-12

    The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

  4. Nuclear reactor control column

    DOE Patents [OSTI]

    Bachovchin, Dennis M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  5. Slurry reactor design studies

    SciTech Connect (OSTI)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. ); Akgerman, A. ); Smith, J.M. )

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  6. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  7. US ITER (International Thermonuclear Experimental Reactor) shield and blanket design activities

    SciTech Connect (OSTI)

    Baker, C.C.

    1988-08-01

    This paper summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. Primary tasks carried out during the past year include design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components, and issues regarding structural materials for an ITER device. The blanket concepts considered are the aqueous/Li salt solution, a water-cooled, solid breeder blanket, a helium-cooled, solid-breeder blanket, a blanket cooled by helium containing lithium-bearing particulates, and a blanket concept based on breeding tritium from He/sup 3/. 1 ref., 2 tabs.

  8. Fast Breeder Reactor studies

    SciTech Connect (OSTI)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  9. CONTROL FOR NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Lichtenberger, H.V.; Cameron, R.A.

    1959-03-31

    S>A control rod operating device in a nuclear reactor of the type in which the control rod is gradually withdrawn from the reactor to a position desired during stable operation is described. The apparatus is comprised essentially of a stop member movable in the direction of withdrawal of the control rod, a follower on the control rod engageable with the stop and means urging the follower against the stop in the direction of withdrawal. A means responsive to disengagement of the follower from the stop is provided for actuating the control rod to return to the reactor shut-down position.

  10. Microfluidic electrochemical reactors

    DOE Patents [OSTI]

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  11. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, Ronald J.; Land, John T.; Misvel, Michael C.

    1994-01-01

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

  12. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, R.J.; Land, J.T.; Misvel, M.C.

    1994-06-07

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

  13. COOLED NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Binner, C.R.; Wilkie, C.B.

    1958-03-18

    This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

  14. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Martin Peng, Y.K.M.

    1985-10-03

    The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

  15. Reactor safety assessment system

    SciTech Connect (OSTI)

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category.

  16. NUCLEAR REACTOR FUEL SYSTEMS

    DOE Patents [OSTI]

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  17. Small Modular Nuclear Reactors: Parametric Modeling of Integrated...

    Office of Environmental Management (EM)

    Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel ... Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel ...

  18. Friction-Induced Fluid Heating in Nanoscale Helium Flows

    SciTech Connect (OSTI)

    Li Zhigang [Department of Mechanical Engineering, Hong Kong University of Science and Technology, Clear Water Bay, Kowloon (Hong Kong)

    2010-05-21

    We investigate the mechanism of friction-induced fluid heating in nanoconfinements. Molecular dynamics simulations are used to study the temperature variations of liquid helium in nanoscale Poiseuille flows. It is found that the fluid heating is dominated by different sources of friction as the external driving force is changed. For small external force, the fluid heating is mainly caused by the internal viscous friction in the fluid. When the external force is large and causes fluid slip at the surfaces of channel walls, the friction at the fluid-solid interface dominates over the internal friction in the fluid and is the major contribution to fluid heating. An asymmetric temperature gradient in the fluid is developed in the case of nonidentical walls and the general temperature gradient may change sign as the dominant heating factor changes from internal to interfacial friction with increasing external force.

  19. Effect of helium on the electronic structure of palladium tritide

    SciTech Connect (OSTI)

    Gupta, R.P.; Gupta, M.

    1998-12-31

    Tritium is usually stored in the form of a metal tritide since it is safe to handle in this form, easily recoverable, and further large quantities of tritium can be stored. However, since tritium is radioactive it decays into {sup 3}He and an electron. Helium recoil energy in this reaction is very small, and not enough to create defects. The authors have performed ab-initio electronic structure calculations that show that in PdT, a considerable amount of {sup 3}He can be accommodated at the octahedral interstitial sites where it is produced. Their calculations also show that the presence of {sup 3}He results in an overall enhancement in the strength of the metal-tritium bonding that leads to the lowering of the plateau pressure. They also find that there is a weakening of the metal-metal bonds due to the repulsive interaction with {sup 3}He.

  20. Options for Cryogenic Load Cooling with Forced Flow Helium Circulation

    SciTech Connect (OSTI)

    Peter Knudsen, Venkatarao Ganni, Roberto Than

    2012-06-01

    Cryogenic pumps designed to circulate super-critical helium are commonly deemed necessary in many super-conducting magnet and other cooling applications. Acknowledging that these pumps are often located at the coldest temperature levels, their use introduces risks associated with the reliability of additional rotating machinery and an additional load on the refrigeration system. However, as it has been successfully demonstrated, this objective can be accomplished without using these pumps by the refrigeration system, resulting in lower system input power and improved reliability to the overall cryogenic system operations. In this paper we examine some trade-offs between using these pumps vs. using the refrigeration system directly with examples of processes that have used these concepts successfully and eliminated using such pumps

  1. Symmetric And Non-Symmetric Muonic Helium Atoms Studies

    SciTech Connect (OSTI)

    Mohammadi, S.

    2011-10-28

    The ground state hyperfine structure and other properties are calculated for muonic helium atoms ({sup 3}He{sup +2}{mu}{sup -}e{sup -} or {sup 4}He{sup +2}{mu}{sup -}e{sup -}) with the use of some local properties of the wave functions in the domains where two particles are close to each other or far away. Simple wave functions incorporating these properties with one variational parameter is developed. The calculated values for hyperfine structure, energy and average interparticle distances in ground state are compared with the correlation function hyper-spherical harmonic method and multibox variational approach. The obtained results are very close to the values calculated by mentioned methods, giving strong indications that the proposed wave functions provide relatively accurate values.

  2. Superconducting cable cooling system by helium gas at two pressures

    DOE Patents [OSTI]

    Dean, John W.

    1977-01-01

    Thermally contacting, oppositely streaming, cryogenic fluid streams in the same enclosure in a closed cycle that changes the fluid from a cool high pressure helium gas to a cooler reduced pressure helium gas in an expander so as to be at different temperature ranges and pressures respectively in go and return legs that are in thermal contact with each other and in thermal contact with a longitudinally extending superconducting transmission line enclosed in the same cable enclosure that insulates the line from the ambient at a temperature T.sub.1. By first circulating the fluid from a refrigerator at one end of the line as a cool gas at a temperature range T.sub.2 to T.sub.3 in the go leg, then circulating the gas through an expander at the other end of the line where the gas becomes a cooler gas at a reduced pressure and at a reduced temperature T.sub.4 and finally by circulating the cooler gas back again to the refrigerator in a return leg at a temperature range T.sub.4 to T.sub.5, while in thermal contact with the gas in the go leg, and in the same enclosure therewith for compression into a higher pressure gas at T.sub.2 in a closed cycle, where T.sub.2 >T.sub.3 and T.sub.5 >T.sub.4, the fluid leaves the enclosure in the go leg as a gas at its coldest point in the go leg, and the temperature distribution is such that the line temperature decreases along its length from the refrigerator due to the cooling from the gas in the return leg.

  3. P Reactor Grouting

    SciTech Connect (OSTI)

    2010-01-01

    Filling the P Reactor with grout. This seals the radioactive material and reduces the environmental footprint left from the Cold War. Project sponsored by the Recovery Act at the Savannah River Site.

  4. Reactor hot spot analysis

    SciTech Connect (OSTI)

    Vilim, R.B.

    1985-08-01

    The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs.

  5. NEUTRONIC REACTOR STRUCTURE

    DOE Patents [OSTI]

    Daniels, F.

    1961-10-24

    A reactor core, comprised of vertical stacks of hexagonal blocks of beryllium oxide having axial cylindrical apertures extending therethrough and cylindrical rods of a sintered mixture of uranium dioxide and beryllium oxide, is described. (AEC)

  6. Molten metal reactors

    DOE Patents [OSTI]

    Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

    2013-11-05

    A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

  7. Future reactor experiments

    SciTech Connect (OSTI)

    Wen, Liangjian

    2015-07-15

    The non-zero neutrino mixing angle θ{sub 13} has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper.

  8. Compact power reactor

    DOE Patents [OSTI]

    Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

  9. K-Reactor readiness

    SciTech Connect (OSTI)

    Rice, P.D.

    1991-12-04

    This document describes some of the more significant accomplishments in the reactor restart program and details the magnitude and extent of the work completed to bring K-Reactor to a state of restart readiness. The discussion of restart achievements is organized into the three major categories of personnel, programs, and plant. Also presented is information on the scope and extent of internal and external oversight of the efforts, as well as some details on the startup plan.

  10. Facile time-of-flight methods for characterizing pulsed superfluid helium droplet beams

    SciTech Connect (OSTI)

    He, Yunteng; Zhang, Jie; Li, Yang; Freund, William M.; Kong, Wei

    2015-08-15

    We present two facile time-of-flight (TOF) methods of detecting superfluid helium droplets and droplets with neutral dopants. Without an electron gun and with only a heated filament and pulsed electrodes, the electron impact ionization TOF mass spectrometer can resolve ionized helium clusters such as He{sub 2}{sup +} and He{sub 4}{sup +}, which are signatures of superfluid helium droplets. Without ionizing any helium atoms, multiphoton non-resonant laser ionization of CCl{sub 4} doped in superfluid helium droplets at 266 nm generates complex cluster ions of dopant fragments with helium atoms, including (He){sub n}C{sup +}, (He){sub n}Cl{sup +}, and (He){sub n}CCl{sup +}. Using both methods, we have characterized our cryogenic pulsed valve—the Even-Lavie valve. We have observed a primary pulse with larger helium droplets traveling at a slower speed and a rebound pulse with smaller droplets at a faster speed. In addition, the pickup efficiency of dopant is higher for the primary pulse when the nozzle temperature is higher than 13 K, and the total time duration of the doped droplet pulse is only on the order of 20 μs. These results stress the importance of fast and easy characterization of the droplet beam for sensitive measurements such as electron diffraction of doped droplets.

  11. Preparation and thermal desorption properties of dc sputtered zirconium-hydrogen-helium thin films

    SciTech Connect (OSTI)

    Wei, Y. C.; Shi, L. Q.; Zhang, L.; He, Z. J.; Zhang, B.; Wang, L. B.

    2008-11-15

    We developed a new approach for preparing hydrogen and helium co-containing zirconium films (Zr-H-He) to simulate aging metal tritides. We also studied the effect of hydrogen on helium behavior, in which we applied direct current magnetron sputtering in a mixture of working gases (helium, argon, and hydrogen). The amount and depth profile of helium and hydrogen trapped in the films were determined using the elastic recoil detection analysis. The microstructure and surface morphology of the Zr-H-He films were studied by x-ray diffraction, transmission electron microscopy, and atomic force microscopy. To investigate the effect of hydrogen on the thermal release behavior of helium in the Zr film, thermal desorption spectrometry (TDS) was used, which revealed a similar desorption behavior to aged tritides. TDS experiments showed that the spectra were constituted by low-temperature peaks around 300 deg. C and high temperature peaks above 750 deg. C. Furthermore, the solid-phase {alpha} to {delta} transformation changed the shapes of the high-temperature peaks related to microstates of helium bubbles and caused the peak with a massive helium release shift toward lower temperature obviously.

  12. F Reactor Inspection

    SciTech Connect (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-10-29

    Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

  13. F Reactor Inspection

    ScienceCinema (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-11-24

    Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

  14. Gas tagging and cover gas combination for nuclear reactor

    DOE Patents [OSTI]

    Gross, Kenny C.; Laug, Matthew T.

    1985-01-01

    The invention discloses the use of stable isotopes of neon and argon, that are grouped in preselected different ratios one to the other and are then sealed as tags in different cladded nuclear fuel elements to be used in a liquid metal fast breeder reactor. Failure of the cladding of any fuel element allows fission gases generated in the reaction and these tag isotopes to escape and to combine with the cover gas held in the reactor over the fuel elements. The isotopes specifically are Ne.sup.20, Ne.sup.21 and Ne.sup.22 of neon and Ar.sup.36, Ar.sup.38 and Ar.sup.40 of argon, and the cover gas is helium. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between approximately 0.degree. and -25.degree. C. operable to remove the fission gases from the cover gas and tags and the second or tag recovery system bed is held between approximately -170.degree. and -185.degree. C. operable to isolate the tags from the cover gas. Spectrometric analysis further is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be specifically determined.

  15. NEUTRONIC REACTOR CONSTRUCTION AND OPERATION

    DOE Patents [OSTI]

    West, J.M.; Weills, J.T.

    1960-03-15

    A method is given for operating a nuclear reactor having a negative coefficient of reactivity to compensate for the change in reactor reactivity due to the burn-up of the xenon peak following start-up of the reactor. When it is desired to start up the reactor within less than 72 hours after shutdown, the temperature of the reactor is lowered prior to start-up, and then gradually raised after start-up.

  16. Effects of helium content of microstructural development in Type 316 stainless steel under neutron irradiation

    SciTech Connect (OSTI)

    Maziasz, P.J.

    1985-11-01

    This work investigated the sensitivity of microstructural evolution, particularly precipitate development, to increased helium content during thermal aging and during neutron irradiation. Helium (110 at. ppM) was cold preinjected into solution annealed (SA) DO-heat type 316 stainess steel (316) via cyclotron irradiation. These specimens were then exposed side by side with uninjected samples. Continuous helium generation was increased considerably relative to EBR-II irradiation by irradiation in HFIR. Data were obtained from quantitative analytical electron microscopy (AEM) in thin foils and on extraction replicas. 480 refs., 86 figs., 19 tabs.

  17. DEUTERIUM, TRITIUM, AND HELIUM DESORPTION FROM AGED TITANIUM TRITIDES. PART I.

    SciTech Connect (OSTI)

    Shanahan, K; Jeffrey Holder, J

    2006-07-10

    Six new samples of tritium-aged bulk titanium have been examined by thermal desorption and isotope exchange chemistry. The discovery of a lower temperature hydrogen desorption state in these materials, previously reported, has been confirmed in one of the new samples. The helium release of the samples shows the more severe effects obtained from longer aging periods, i.e. higher initial He/M ratios. Several of the more aged samples were spontaneously releasing helium. Part I will discuss the new results on the new lower temperature hydrogen desorption state found in one more extensively studied sample. Part II will discuss the hydrogen/helium release behavior of the remaining samples.

  18. DEUTERIUM, TRITIUM, AND HELIUM DESORPTION FROM AGED TITANIUM TRITIDES. PART II.

    SciTech Connect (OSTI)

    Shanahan, K; Jeffrey Holder, J

    2006-08-17

    Six new samples of tritium-aged bulk titanium have been examined by thermal desorption and isotope exchange chemistry. The discovery of a lower temperature hydrogen desorption state in these materials, previously reported, has been confirmed in one of the new samples. The helium release of the samples shows the more severe effects obtained from longer aging periods, i.e. higher initial He/M ratios. Several of the more aged samples were spontaneously releasing helium. Part I discussed the new results on the new lower temperature hydrogen desorption state found in one more extensively studied sample. Part II will discuss the hydrogen/helium release behavior of the remaining samples.

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

    SciTech Connect (OSTI)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  20. Helium bubble evolution in a Zr–Sn–Nb–Fe–Cr alloy during post-annealing: An in-situ investigation

    SciTech Connect (OSTI)

    Shen, H.H.; Peng, S.M.; Chen, B.; Naab, F.N.; Sun, G.A.; Zhou, W.; Xiang, X.; Sun, K.; Zu, X.T.

    2015-09-15

    The formation of helium bubbles is considered to be detrimental to the mechanical performance of the nuclear materials. The growth behaviors of helium bubbles in a helium ion implanted Zr–Sn–Nb–Fe–Cr alloy with respect to the helium fluence and subsequently annealing procedure were investigated by in-situ transmission electron microscopy. In the as-implanted sample, the measured size distributions of the helium bubbles are consistent with the simulated helium concentrations. Moreover, the mean size of the helium bubbles increases with the increase of the irradiation temperatures and the helium fluence. The in-situ heating study performed in a transmission electron microscope indicates that the mean size of the helium bubbles increase slowly below 923 K and dramatically above 923 K. The coarsening mechanism of the helium bubbles in the alloy is suggested based on the study. - Highlights: • Helium bubble growth in zirconium with annealing was in-situ investigated in TEM. • The mean helium bubble size increase with helium fluence and annealing temperature. • Helium bubble size distribution is same as that of helium concentration by SRIM. • Mean bubble size increases slowly and quickly with temperature below and above 923 K. • The growth mechanism of the helium bubbles in Zr alloy has been discussed.

  1. REACTOR GROUT THERMAL PROPERTIES

    SciTech Connect (OSTI)

    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

    2011-01-28

    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

  2. Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

    SciTech Connect (OSTI)

    Klein, Andrew; Matthews, Topher; Lenhof, Renae; Deason, Wesley; Harter, Jackson

    2015-01-16

    Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

  3. Investigations of the Application of CFD to Flow Expected in the Lower Plenum of the Prismatic VHTR

    SciTech Connect (OSTI)

    Richard W.Johnson; Tara Gallaway; Donna P. Guillen

    2006-09-01

    The Generation IV (Gen IV) very high temperature reactor (VHTR) will either be a prismatic (block) or pebble bed design. However, a prismatic VHTR reference design, based on the General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) [General Atomics, 1996] has been developed for preliminary analysis purposes [MacDonald, et al., 2003]. Numerical simulation studies reported herein are based on this reference design. In the lower plenum of the prismatic reference design, the flow will be introduced by dozens of turbulent jets from the core above. The jet flow will encounter rows of columns that support the core. The flow from the core will have to turn ninety degrees and flow toward the exit duct as it passed through the forest of support columns. Due to the radial variation of the power density in the core, the jets will be at various temperatures at the inlet to the lower plenum. This presents some concerns, including that local hot spots may occur in the lower plenum. This may have a deleterious effect on the materials present as well as cause a variation in temperature to be present as the flow enters the power conversion system machinery, which could cause problems with the operation of the machinery. In the past, systems analysis codes have been used to model flow in nuclear reactor systems. It is recognized, however, that such codes are not capable of modeling the local physics of the flow to be able to analyze for local mixing and temperature variations. This has led to the determination that computational fluid dynamic (CFD) codes be used, which are generally regarded as having the capability of accurately simulating local flow physics. Accurate flow modeling involves determining appropriate modeling strategies needed to obtain accurate analyses. These include determining the fineness of the grid needed, the required iterative convergence tolerance, which numerical discretization method to use, and which turbulence model and wall treatment should be

  4. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    SciTech Connect (OSTI)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  5. Advanced Reactor Concepts Technical Review Panel Report | Department...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    The concepts included five fast reactors and three thermal reactors. As to reactor coolants, there were three sodium-cooled reactors, two gas-cooled reactors, one light ...

  6. REACTOR AND NOVEL METHOD

    DOE Patents [OSTI]

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

    1958-06-24

    A nuclear reactor of the type which uses a liquid fuel and a method of controlling such a reactor are described. The reactor is comprised essentially of a tank for containing the liquid fuel such as a slurry of discrete particles of fissionnble material suspended in a heavy water moderator, and a control means in the form of a disc of neutron absorbirg material disposed below the top surface of the slurry and parallel thereto. The diameter of the disc is slightly smaller than the diameter of the tank and the disc is perforated to permit a flow of the slurry therethrough. The function of the disc is to divide the body of slurry into two separate portions, the lower portion being of a critical size to sustain a nuclear chain reaction and the upper portion between the top surface of the slurry and the top surface of the disc being of a non-critical size. The method of operation is to raise the disc in the reactor until the lower portion of the slurry has reached a critical size when it is desired to initiate the reaction, and to lower the disc in the reactor to reduce the size of the lower active portion the slurry to below criticality when it is desired to stop the reaction.

  7. Light Water Reactor Sustainability Technical Documents | Department...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    in Light Water Reactors: Life After 60 Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high...

  8. Zero Power Reactor simulation | Argonne National Laboratory

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Zero Power Reactor simulation Share Description Ever wanted to see a nuclear reactor core in action? Here's a detailed simulation of the Zero Power Reactor experiment, run by ...

  9. Note: Control of liquid helium supply to cryopanels of Kolkata superconducting cyclotron

    SciTech Connect (OSTI)

    Bhattacharyya, T. K. Pal, G.

    2015-02-15

    The Kolkata superconducting cyclotron utilises liquid helium to cool the main magnet niobium-titanium (NbTi) coil and the cryopanels. Three liquid helium cooled cryopanels, placed inside the dees of the radio-frequency system, maintain the high vacuum in the acceleration region of the superconducting cyclotron. The small cryostat placed inside the cryogenic distribution manifold located at the basement of the superconducting cyclotron building supplies liquid helium in parallel branches to three cold heads, used for cooling their associated cryopanels. The level in the cryostat has to be maintained at an optimum value to ensure uninterrupted flow of liquid helium to these three cold heads. This paper describes the transfer function of the overall system, its tuning parameters, and discusses the actual control of cryostat level by using these parameters.

  10. Research and development of a helium-4 based solar neutrino detector

    SciTech Connect (OSTI)

    Lanou, R.E.; Maris, H.J.; Seidel, G.M.

    1990-12-01

    We report on work accomplished in the first 30 months of a research and development program to investigate the feasibility of a new technique to detect solar neutrinos in superfluid helium. Accomplishments include the successful completion of design, construction and operation of the entire cryogenic, mechanical and electronic apparatus. During the last several months we have begun a series of experiments in superfluid helium to test the method. Experimental results include the first observation of the combined physical processes essential to the detection technique: ballistic roton generation by energetic charged particles, quantum evaporation of helium at a free surface and bolometric detection of the evaporated helium by physisorption on a cold silicon wafer. Additional results are also presented.

  11. Helium release rates and ODH calculations from RHIC magnet cooling line failure

    SciTech Connect (OSTI)

    Liaw, C.J.; Than, Y.; Tuozzolo, J.

    2011-03-28

    A catastrophic failure of the magnet cooling lines, similar to the LHC superconducting bus failure incident, could discharge cold helium into the RHIC tunnel and cause an Oxygen Deficiency Hazard (ODH) problem. A SINDA/FLUINT{reg_sign} model, which simulated the 4.5K/4 atm helium flowing through the magnet cooling system distribution lines, then through a line break into the insulating vacuum volumes and discharging via the reliefs into the RHIC tunnel, had been developed. Arc flash energy deposition and heat load from the ambient temperature cryostat surfaces are included in the simulations. Three typical areas: the sextant arc, the Triplet/DX/D0 magnets, and the injection area, had been analyzed. Results, including helium discharge rates, helium inventory loss, and the resulting oxygen concentration in the RHIC tunnel area, are reported. Good agreement had been achieved when comparing the simulation results, a RHIC sector depressurization test measurement, and some simple analytical calculations.

  12. Evaluation of helium impurity impacts on Spent Nuclear Fuel project processes (OCRWM)

    SciTech Connect (OSTI)

    SHERRELL, D.L.

    1999-09-21

    This document identifies the types and quantities of impurities that may be present within helium that is introduced into multi-canister overpacks (MCO)s by various SNF Project facilities, including, but not limited to the Cold Vacuum Drying (CVD) Facility (CVDF). It then evaluates possible impacts of worst case impurity inventories on MCO drying, transportation, and storage processes. Based on the evaluation results, this document: (1) concludes that the SNF Project helium procurement specification can be a factor-of-ten less restrictive than a typical vendor's standard offering (99.96% pure versus the vendor's 99.997% pure standard offering); (2) concludes that the CVDF's current 99.5% purity requirement is adequate to control the quality of the helium that is delivered to the MCO by the plant's helium distribution system; and (3) recommends specific impurity limits for both of the above cases.

  13. Intermediate Heat Transfer Loop Study for High Temperature Gas-Cooled Reactor

    SciTech Connect (OSTI)

    C. H. Oh; C. Davis; S. Sherman

    2008-08-01

    A number of possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermal-hydraulic and cycleefficiency evaluations of the different configurations and coolants. The thermal-hydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. This paper also includes a portion of stress analyses performed on pipe configurations.

  14. Methanation assembly using multiple reactors

    DOE Patents [OSTI]

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  15. Medium Power Lead Alloy Fast Reactor Balance of Plant Options

    SciTech Connect (OSTI)

    Vaclav Dosta; Pavel Hejzlar; Neil E. Todreas; Jacopo Buongiorno

    2004-09-01

    Proper selection of the power conversion cycle is a very important step in the design of a nuclear reactor. Due to the higher core outlet temperature (~550°C) compared to that of light water reactors (~300°C), a wide portfolio of power cycles is available for the lead alloy fast reactor (LFR). Comparison of the following cycles for the LFR was performed: superheated steam (direct and indirect), supercritical steam, helium Brayton, and supercritical CO2 (S-CO2) recompression. Heat transfer from primary to secondary coolant was first analyzed and then the steam generators or heat exchangers were designed. The direct generation of steam in the lead alloy coolant was also evaluated. The resulting temperatures of the secondary fluids are in the range of 530-545°C, dictated by the fixed space available for the heat exchangers in the reactor vessel. For the direct steam generation situation, the temperature is 312°C. Optimization of each power cycle was carried out, yielding net plant efficiency of around 40% for the superheated steam cycle while the supercritical steam and S-CO2 cycles achieved net plant efficiency of 41%. The cycles were then compared based on their net plant efficiency and potential for low capital cost. The superheated steam cycle is a very good candidate cycle given its reasonably high net plant efficiency and ease of implementation based on the extensive knowledge and operating experience with this cycle. Although the supercritical steam cycle net plant efficiency is slightly better than that of the superheated steam cycle, its high complexity and high pressure result in higher capital cost, negatively affecting plant economics. The helium Brayton cycle achieves low net plant efficiency due to the low lead alloy core outlet temperature, and therefore, even though it is a simpler cycle than the steam cycles, its performance is mediocre in this application. The prime candidate, however, appears to be the S-CO2 recompression cycle, because it

  16. Exploring the isopycnal mixing and helium-heat paradoxes in a suite of Earth System Models

    DOE PAGES-Beta [OSTI]

    Gnanadesikan, A.; Abernathey, R.; Pradal, M.-A.

    2014-11-20

    This paper uses a suite of Earth System models which simulate the distribution of He isotopes and radiocarbon to examine two paradoxes in Earth science. The helium-heat paradox refers to the fact that helium emissions to the deep ocean are far lower than would be expected given the rate of geothermal heating, since both are thought to be the result of radioactive decay in the earth's interior. The isopycnal mixing paradox comes from the fact that many theoretical parameterizations of the isopycnal mixing coefficient ARedi that link it to baroclinic instability project it to be small (of order a fewmore » hundred m2 s−1) in the ocean interior away from boundary currents. However, direct observations using tracers and floats (largely in the upper ocean) suggest that values of this coefficient are an order of magnitude higher. Because helium isotopes equilibrate rapidly with the atmosphere, but radiocarbon equilibrates slowly, it might be thought that resolving the isopycnal mixing paradox in favor of the higher observational estimates of ARedi might also solve the helium paradox. In this paper we show that this is not the case. In a suite of models with different spatially constant and spatially varying values of ARedi the distribution of radiocarbon and helium isotopes is sensitive to the value of ARedi. However, away from strong helium sources in the Southeast Pacific, the relationship between the two is not sensitive, indicating that large-scale advection is the limiting process for removing helium and radiocarbon from the deep ocean. The helium isotopes, in turn, suggest a higher value of ARedi in the deep ocean than is seen in theoretical parameterizations based on baroclinic growth rates. We argue that a key part of resolving the isopycnal mixing paradox is to abandon the idea that ARedi has a direct relationship to local baroclinic instability and to the so called "thickness" mixing coefficient AGM.« less

  17. Radon and helium in soil gases in the Phlegraean Fields, central Italy

    SciTech Connect (OSTI)

    Lombardi, S. ); Reimer, G.M. )

    1990-05-01

    The distribution and migration of radon and helium soil-gas concentrations in the Phlegraean Fields, Italy, are controlled by the tectonic features of the area. Radon is supplied from surficial sources and helium has both surficial and deep origins. There is no direct correlation between the two noble gases on a point-to-point basis but the areal distribution of both gases is similar, suggesting that the distribution is controlled primarily by fractures and movement of geothermal fluids.

  18. Breazeale Reactor Modernization Program

    SciTech Connect (OSTI)

    Davison, C. C.

    2003-04-16

    The Penn State Breazeale Nuclear Reactor is the longest operating licensed research reactor in the nation. The facility has played a key role in educating scientists, engineers and in providing facilities and services to researchers in many different disciplines. In order to remain a viable and effective research and educational institution, a multi-phase modernization project was proposed. Phase I was the replacement of the 25-year old reactor control and safety system along with associated wiring and hardware. This phase was fully funded by non-federal funds. Tasks identified in Phases II-V expand upon and complement the work done in Phase I to strategically implement state-of-the-art technologies focusing on identified national needs and priorities of the future.

  19. Reactor for exothermic reactions

    DOE Patents [OSTI]

    Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

    1993-03-02

    A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  20. Reactor for exothermic reactions

    DOE Patents [OSTI]

    Smith, Jr., Lawrence A. (Bellaire, TX); Hearn, Dennis (Houston, TX); Jones, Jr., Edward M. (Friendswood, TX)

    1993-01-01

    A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120.degree. to 300.degree. F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  1. Dynamic bed reactor

    DOE Patents [OSTI]

    Stormo, Keith E. (Moscow, ID)

    1996-07-02

    A dynamic bed reactor is disclosed in which a compressible open cell foam matrix is periodically compressed and expanded to move a liquid or fluid through the matrix. In preferred embodiments, the matrix contains an active material such as an enzyme, biological cell, chelating agent, oligonucleotide, adsorbent or other material that acts upon the liquid or fluid passing through the matrix. The active material may be physically immobilized in the matrix, or attached by covalent or ionic bonds. Microbeads, substantially all of which have diameters less than 50 microns, can be used to immobilize the active material in the matrix and further improve reactor efficiency. A particularly preferred matrix is made of open cell polyurethane foam, which adsorbs pollutants such as polychlorophenol or o-nitrophenol. The reactors of the present invention allow unidirectional non-laminar flow through the matrix, and promote intimate exposure of liquid reactants to active agents such as microorganisms immobilized in the matrix.

  2. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  3. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, Anstein; Lazarus, Jonathan D.

    1987-01-01

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

  4. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, Ernest

    1986-01-01

    A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

  5. A NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Luebke, E.A.; Vandenberg, L.B.

    1959-09-01

    A nuclear reactor for producing thermoelectric power is described. The reactor core comprises a series of thermoelectric assemblies, each assembly including fissionable fuel as an active element to form a hot junction and a thermocouple. The assemblies are disposed parallel to each other to form spaces and means are included for Introducing an electrically conductive coolant between the assemblies to form cold junctions of the thermocouples. An electromotive force is developed across the entire series of the thermoelectric assemblies due to fission heat generated in the fuel causing a current to flow perpendicular to the flow of coolant and is distributed to a load outside of the reactor by means of bus bars electrically connected to the outermost thermoelectric assembly.

  6. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-08-01

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  7. Thermal analysis for fuel handling system for sodium cooled reactor considering minor actinide-bearing metal fuel.

    SciTech Connect (OSTI)

    Chikazawa, Y.; Grandy, C.; Nuclear Engineering Division

    2009-03-01

    The Advanced Burner Reactor (ABR) is one of the components of the Global Nuclear Energy Partnership (GNEP) used to close the fuel cycle. ABR is a sodium-cooled fast reactor that is used to consume transuranic elements resulting from the reprocessing of light water reactor spent nuclear fuel. ABR-1000 [1000 MW(thermal)] is a fast reactor concept created at Argonne National Laboratory to be used as a reference concept for various future trade-offs. ABR-1000 meets the GNEP goals although it uses what is considered base sodium fast reactor technology for its systems and components. One of the considerations of any fast reactor plant concept is the ability to perform fuel-handling operations with new and spent fast reactor fuel. The transmutation fuel proposed as the ABR fuel has a very little experience base, and thus, this paper investigates a fuel-handling concept and potential issues of handling fast reactor fuel containing minor actinides. In this study, two thermal analyses supporting a conceptual design study on the ABR-1000 fuel-handling system were carried out. One analysis investigated passive dry spent fuel storage, and the other analysis investigated a fresh fuel shipping cask. Passive dry storage can be made suitable for the ABR-1000 spent fuel storage with sodium-bonded metal fuel. The thermal analysis shows that spent fast reactor fuel with a decay heat of 2 kW or less can be stored passively in a helium atmosphere. The 2-kW value seems to be a reasonable and practical level, and a combination of reasonably-sized in-sodium storage followed by passive dry storage could be a candidate for spent fuel storage for the next-generation sodium-cooled reactor with sodium-bonded metal fuel. Requirements for the shipping casks for minor actinide-bearing fuel with a high decay heat level are also discussed in this paper. The shipping cask for fresh sodium-cooled-reactor fuel should be a dry type to reduce the reaction between residual moisture on fresh fuel and the

  8. Experimental study of forced convection heat transfer during upward and downward flow of helium at high pressure and high temperature

    SciTech Connect (OSTI)

    Francisco Valentin; Narbeh Artoun; Masahiro Kawaji; Donald M. McEligot

    2015-08-01

    Fundamental high pressure/high temperature forced convection experiments have been conducted in support of the development of a Very High Temperature Reactor (VHTR) with a prismatic core. The experiments utilize a high temperature/high pressure gas flow test facility constructed for forced convection and natural circulation experiments. The test section has a single 16.8 mm ID flow channel in a 2.7 m long, 108 mm OD graphite column with four 2.3kW electric heater rods placed symmetrically around the flow channel. This experimental study presents the role of buoyancy forces in enhancing or reducing convection heat transfer for helium at high pressures up to 70 bar and high temperatures up to 873 degrees K. Wall temperatures have been compared among 10 cases covering the inlet Re numbers ranging from 500 to 3,000. Downward flows display higher and lower wall temperatures in the upstream and downstream regions, respectively, than the upward flow cases due to the influence of buoyancy forces. In the entrance region, convection heat transfer is reduced due to buoyancy leading to higher wall temperatures, while in the downstream region, buoyancyinduced mixing causes higher convection heat transfer and lower wall temperatures. However, their influences are reduced as the Reynolds number increases. This experimental study is of specific interest to VHTR design and validation of safety analysis codes.

  9. Performance Testing of Jefferson Lab 12 GeV Helium Screw Compressors

    DOE PAGES-Beta [OSTI]

    Knudsen, P.; Ganni, V.; Dixon, K.; Norton, R.; Creel, J.

    2015-08-10

    Oil injected screw compressors have essentially superseded all other types of compressors in modern helium refrigeration systems due to their large displacement capacity, reliability, minimal vibration, and capability of handling helium's high heat of compression. At the present state of compressor system designs for helium refrigeration systems, typically two-thirds of the lost input power is due to the compression system. It is important to understand the isothermal and volumetric efficiencies of these machines to help properly design the compression system to match the refrigeration process. It is also important to identify those primary compressor skid exergetic loss mechanisms which maymore » be reduced, thereby offering the possibility of significantly reducing the input power to helium refrigeration processes which are extremely energy intensive. This paper summarizes the results collected during the commissioning of the new compressor system for Jefferson Lab's (JLab's) 12 GeV upgrade. The compressor skid packages were designed by JLab and built to print by industry. They incorporate a number of modifications not typical of helium screw compressor packages and most importantly allow a very wide range of operation so that JLab's patented Floating Pressure Process can be fully utilized. This paper also summarizes key features of the skid design that allow this process and facilitate the maintenance and reliability of these helium compressor systems.« less

  10. The initiation and propagation of helium detonations in white dwarf envelopes

    SciTech Connect (OSTI)

    Shen, Ken J. [Department of Astronomy and Theoretical Astrophysics Center, University of California, Berkeley, CA 94720 (United States); Moore, Kevin, E-mail: kenshen@astro.berkeley.edu [Department of Applied Mathematics and Statistics, University of California, Santa Cruz, CA 95064 (United States)

    2014-12-10

    Detonations in helium-rich envelopes surrounding white dwarfs have garnered attention as triggers of faint thermonuclear '.Ia' supernovae and double detonation Type Ia supernovae. However, recent studies have found that the minimum size of a hotspot that can lead to a helium detonation is comparable to, or even larger than, the white dwarf's pressure scale height, casting doubt on the successful ignition of helium detonations in these systems. In this paper, we examine the previously neglected effects of C/O pollution and a full nuclear reaction network, and we consider hotspots with spatially constant pressure in addition to constant density hotspots. We find that the inclusion of these effects significantly decreases the minimum hotspot size for helium-rich detonation ignition, making detonations far more plausible during turbulent shell convection or during double white dwarf mergers. The increase in burning rate also decreases the minimum shell mass in which a helium detonation can successfully propagate and alters the composition of the shell's burning products. The ashes of these low-mass shells consist primarily of silicon, calcium, and unburned helium and metals and may explain the high-velocity spectral features observed in most Type Ia supernovae.

  11. Performance Testing of Jefferson Lab 12 GeV Helium Screw Compressors

    SciTech Connect (OSTI)

    Knudsen, P.; Ganni, V.; Dixon, K.; Norton, R.; Creel, J.

    2015-08-10

    Oil injected screw compressors have essentially superseded all other types of compressors in modern helium refrigeration systems due to their large displacement capacity, reliability, minimal vibration, and capability of handling helium's high heat of compression. At the present state of compressor system designs for helium refrigeration systems, typically two-thirds of the lost input power is due to the compression system. It is important to understand the isothermal and volumetric efficiencies of these machines to help properly design the compression system to match the refrigeration process. It is also important to identify those primary compressor skid exergetic loss mechanisms which may be reduced, thereby offering the possibility of significantly reducing the input power to helium refrigeration processes which are extremely energy intensive. This paper summarizes the results collected during the commissioning of the new compressor system for Jefferson Lab's (JLab's) 12 GeV upgrade. The compressor skid packages were designed by JLab and built to print by industry. They incorporate a number of modifications not typical of helium screw compressor packages and most importantly allow a very wide range of operation so that JLab's patented Floating Pressure Process can be fully utilized. This paper also summarizes key features of the skid design that allow this process and facilitate the maintenance and reliability of these helium compressor systems.

  12. Fast quench reactor method

    DOE Patents [OSTI]

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.; Berry, Ray A.

    1999-01-01

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

  13. THERMAL NUCLEAR REACTOR

    DOE Patents [OSTI]

    Fenning, F.W.; Jackson, R.F.

    1957-09-24

    Nuclear reactors of the graphite moderated air cooled type in which canned slugs or rods of fissile material are employed are discussed. Such a reactor may be provided with a means for detecting dust particles in the exhausted air. The means employed are lengths of dust absorbent cord suspended in vertical holes in the shielding structure above each vertical coolant flow channel to hang in the path of the cooling air issuing from the channels, and associated spindles and drive motors for hauling the cords past detectors, such as Geiger counters, for inspecting the cords periodically. This design also enables detecting the individual channel in which a fault condition may have occurred.

  14. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, Daniel L.

    1988-01-01

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.

  15. NEUTRONIC REACTOR CONTROL ELEMENT

    DOE Patents [OSTI]

    Newson, H.W.

    1960-09-13

    A novel composite neutronic reactor control element is offered. The element comprises a multiplicity of sections arranged in end-to-end relationship, each of the sections having a markedly different neutron-reactive characteristic. For example, a three-section control element could contain absorber, moderator, and fuel sections. By moving such an element longitudinally through a reactor core, reactivity is decreased by the absorber, increased slightly by the moderator, or increased substantially by the fuel. Thus, control over a wide reactivity range is provided.

  16. Perspectives on reactor safety

    SciTech Connect (OSTI)

    Haskin, F.E.; Camp, A.L.

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  17. Repair of hydroprocessing reactors

    SciTech Connect (OSTI)

    Niccolls, E.H.; Imgram, A.G.; Bagdasarian, A.J.

    1995-12-01

    In this paper the authors very briefly review the types of damage observed and repairs performed in hydroprocessing reactors. The authors will summarize the fundamental damage mechanisms, and where in the reactors they have needed to make repairs over the years. The era from the 1960`s through about 1990 will be briefly discussed. They describe in more detail repairs undertaken in the 1990`s. Finally, they will note important technical issues their industry faces regarding repairs, and the long term reliable operation of these vessels.

  18. Nuclear reactor apparatus

    DOE Patents [OSTI]

    Wade, Elman E.

    1978-01-01

    A lifting, rotating and sealing apparatus for nuclear reactors utilizing rotating plugs above the nuclear reactor core. This apparatus permits rotation of the plugs to provide under the plug refueling of a nuclear core. It also provides a means by which positive top core holddown can be utilized. Both of these operations are accomplished by means of the apparatus lifting the top core holddown structure off the nuclear core while stationary, and maintaining this structure in its elevated position during plug rotation. During both of these operations, the interface between the rotating member and its supporting member is sealingly maintained.

  19. Plug Flow Reactor Simulator

    Energy Science and Technology Software Center (OSTI)

    1996-07-30

    PLUG is a computer program that solves the coupled steady state continuity, momentum, energy, and species balance equations for a plug flow reactor. Both homogeneous (gas-phase) and heterogenous (surface) reactions can be accommodated. The reactor may be either isothermal or adiabatic or may have a specified axial temperature or heat flux profile; alternatively, an ambient temperature and an overall heat-transfer coefficient can be specified. The crosssectional area and surface area may vary with axial position,more » and viscous drag is included. Ideal gas behavior and surface site conservation are assumed.« less

  20. Fast quench reactor method

    DOE Patents [OSTI]

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.; Berry, R.A.

    1999-08-10

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream. 8 figs.

  1. MEANS FOR SHIELDING REACTORS

    DOE Patents [OSTI]

    Garrison, W.M.; McClinton, L.T.; Burton, M.

    1959-03-10

    A reactor of the heterageneous, heavy water moderated type is described. The reactor is comprised of a plurality of vertically disposed fuel element tubes extending through a tank of heavy water moderator and adapted to accommodate a flow of coolant water in contact with the fuel elements. A tank containing outgoing coolant water is disposed above the core to function is a radiation shield. Unsaturated liquid hydrocarbon is floated on top of the water in the shield tank to reduce to a minimum the possibility of the occurrence of explosive gaseous mixtures resulting from the neutron bombardment of the water in the shield tank.

  2. Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor

    Office of Environmental Management (EM)

    removed from Hanford's 300 Area | Department of Energy Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed from Hanford's 300 Area Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed from Hanford's 300 Area January 22, 2014 - 12:00pm Addthis Media Contacts Cameron Hardy, DOE 509-376-5365 Cameron.Hardy@re.doe.gov Mark McKenna, Washington Closure 509-372-9032 media@wch-rcc.com RICHLAND, WA - Hanford's River Corridor contractor, Washington

  3. Atomic-scale mechanisms of helium bubble hardening in iron

    DOE PAGES-Beta [OSTI]

    Osetskiy, Yury N.; Stoller, Roger E.

    2015-06-03

    Generation of helium due to (n,α) transmutation reactions changes the response of structural materials to neutron irradiation. The whole process of radiation damage evolution is affected by He accumulation and leads to significant changes in the material s properties. A population of nanometric He-filled bubbles affects mechanical properties and the impact can be quite significant because of their high density. Understanding how these basic mechanisms affect mechanical properties is necessary for predicting radiation effects. In this paper we present an extensive study of the interactions between a moving edge dislocation and bubbles using atomic-scale modeling. We focus on the effectmore » of He bubble size and He concentration inside bubbles. Thus, we found that ability of bubbles to act as an obstacle to dislocation motion is close to that of voids when the He-to-vacancy ratio is in the range from 0 to 1. A few simulations made at higher He contents demonstrated that the interaction mechanism is changed for over-pressurized bubbles and they become weaker obstacles. The results are discussed in light of post-irradiation materials testing.« less

  4. Atomic-scale mechanisms of helium bubble hardening in iron

    SciTech Connect (OSTI)

    Osetskiy, Yury N.; Stoller, Roger E.

    2015-06-03

    Generation of helium due to (n,α) transmutation reactions changes the response of structural materials to neutron irradiation. The whole process of radiation damage evolution is affected by He accumulation and leads to significant changes in the material s properties. A population of nanometric He-filled bubbles affects mechanical properties and the impact can be quite significant because of their high density. Understanding how these basic mechanisms affect mechanical properties is necessary for predicting radiation effects. In this paper we present an extensive study of the interactions between a moving edge dislocation and bubbles using atomic-scale modeling. We focus on the effect of He bubble size and He concentration inside bubbles. Thus, we found that ability of bubbles to act as an obstacle to dislocation motion is close to that of voids when the He-to-vacancy ratio is in the range from 0 to 1. A few simulations made at higher He contents demonstrated that the interaction mechanism is changed for over-pressurized bubbles and they become weaker obstacles. The results are discussed in light of post-irradiation materials testing.

  5. Performance Characterization of the Production Facility Prototype Helium Flow System

    SciTech Connect (OSTI)

    Woloshun, Keith Albert; Dale, Gregory E.; Dalmas, Dale Allen; Romero, Frank Patrick

    2015-12-16

    The roots blower in use at ANL for in-beam experiments and also at LANL for flow tests was sized for 12 mm diameter disks and significantly less beam heating. Currently, the disks are 29 mm in diameter, with a 12 mm FWHM Gaussian beam spot at 42 MeV and 2.86 μA on each side of the target, 5.72 μA total. The target design itself is reported elsewhere. With the increased beam heating, the helium flow requirement increased so that a larger blower was need for a mass flow rate of 400 g/s at 2.76 MPa (400 psig). An Aerzen GM 12.4 blower was selected, and is currently being installed at the LANL facility for target and component flow testing. This report describes this blower/motor/pressure vessel package and the status of the facility preparations. Blower performance (mass flow rate as a function of loop pressure drop) was measured at 4 blower speeds. Results are reported below.

  6. B Reactor | Department of Energy

    Energy.gov (indexed) [DOE]

    boomtown, with the population reaching 50,000 by summer 1944. Similar to the X-10 Graphite Reactor at Oak Ridge in terms of loading and unloading fuel, the B Reactor was built...

  7. Reactor operation environmental information document

    SciTech Connect (OSTI)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  8. F Reactor Area Cleanup Complete

    Energy.gov [DOE]

    RICHLAND, Wash. – U.S. Department of Energy (DOE) contractors have cleaned up the F Reactor Area, the first reactor area at the Hanford Site in southeastern Washington state to be fully remediated.

  9. Hallam, Nebraska, Decommissioned Reactor Site

    Office of Legacy Management (LM)

    Fact Sheet D&D D&D Hallam, Nebraska, Decommissioned Reactor Site This fact sheet provides information about the Hallam, Nebraska, Decommissioned Reactor Site. This site is managed by the U.S. Department of Energy Office of Legacy Management under the Defense Decontamination and Decommissioning (D&D) Program. Location of the Hallam Decommissioned Reactor Site Description and History The Hallam decommissioned reactor site is in southeastern Nebraska, approximately 19 miles south of

  10. Reactor operation safety information document

    SciTech Connect (OSTI)

    Not Available

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  11. NETL - Chemical Looping Reactor

    SciTech Connect (OSTI)

    2013-07-24

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  12. JACKETED REACTOR FUEL ELEMENT

    DOE Patents [OSTI]

    Smith, K.F.; Van Thyne, R.J.

    1958-12-01

    A fuel element is described for fast reactors comprised of a core of uranium metal containing material and a jacket around the core, the jacket consisting of from 2.5 to 15 percent of titanium, from 1 to 5 percent of niobium, and from 80 to 96.5 percent of vanadium.

  13. Fossil fuel furnace reactor

    DOE Patents [OSTI]

    Parkinson, William J.

    1987-01-01

    A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

  14. Thermal Reactor Safety

    SciTech Connect (OSTI)

    Not Available

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  15. REACTOR UNLOADING MEANS

    DOE Patents [OSTI]

    Cooper, C.M.

    1957-08-20

    A means for remotely unloading irradiated fuel slugs from a neutronic reactor core and conveying them to a remote storage tank is reported. The means shown is specifically adapted for use with a reactor core wherein the fuel slugs are slidably held in end to end abutting relationship in the horizontal coolant flow tubes, the slugs being spaced from tae internal walls of the tubes to permit continuous circulation of coolant water therethrough. A remotely operated plunger at the charging ends of the tubes is used to push the slugs through the tubes and out the discharge ends into a special slug valve which transfers the slug to a conveying tube leading into a storage tank. Water under pressure is forced through the conveying tube to circulate around the slug to cool it and also to force the slug through the conveving tube into the storage tank. The slug valve and conveying tube are shielded to prevent amy harmful effects caused by the radioactive slug in its travel from the reactor to the storage tank. With the disclosed apparatus, all the slugs in the reactor core can be conveyed to the storage tank shortly after shutdown by remotely located operating personnel.

  16. Nuclear reactor building

    DOE Patents [OSTI]

    Gou, P.F.; Townsend, H.E.; Barbanti, G.

    1994-04-05

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

  17. NEUTRONIC REACTOR SHIELD

    DOE Patents [OSTI]

    Fermi, E.; Zinn, W.H.

    1957-09-24

    The reactor radiation shield material is comprised of alternate layers of iron-containing material and compressed cellulosic material, such as masonite. The shielding material may be prefabricated in the form of blocks, which can be stacked together in ary desired fashion to form an effective shield.

  18. Cermet fuel reactors

    SciTech Connect (OSTI)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  19. NEUTRONIC REACTOR FUEL PUMP

    DOE Patents [OSTI]

    Cobb, W.G.

    1959-06-01

    A reactor fuel pump is described which offers long life, low susceptibility to radiation damage, and gaseous fission product removal. An inert-gas lubricated bearing supports a journal on one end of the drive shsft. The other end has an impeller and expansion chamber which effect pumping and gas- liquid separation. (T.R.H.)

  20. SYSTEM FOR UNLOADING REACTORS

    DOE Patents [OSTI]

    Rand, A.C. Jr.

    1961-05-01

    An unloading device for individual vertical fuel channels in a nuclear reactor is shown. The channels are arranged in parallel rows and underneath each is a separate supporting block on which the fuel in the channel rests. The blocks are raounted in contiguous rows on an array of parallel pairs of tracks over the bottom of the reactor. Oblong hollows in the blocks form a continuous passageway through the middle of the row of blocks on each pair of tracks. At the end of each passageway is a horizontal grappling rod with a T- or L extension at the end next to the reactor of a length to permit it to pass through the oblong passageway in one position, but when rotated ninety degrees the head will strike one of the longer sides of the oblong hollow of one of the blocks. The grappling rod is actuated by a controllable reciprocating and rotating device which extends it beyond any individual block desired, rotates it and retracts it far enough to permit the fuel in the vertical channel above the block to fall into a handling tank below the reactor.

  1. Reactor component automatic grapple

    DOE Patents [OSTI]

    Greenaway, Paul R.

    1982-01-01

    A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment.

  2. NUCLEAR REACTOR FUEL ELEMENT

    DOE Patents [OSTI]

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

    1963-06-11

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

  3. NETL - Chemical Looping Reactor

    ScienceCinema (OSTI)

    None

    2014-06-26

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  4. Nuclear reactor building

    DOE Patents [OSTI]

    Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  5. NEUTRONIC REACTOR CONTROL ELEMENT

    DOE Patents [OSTI]

    Beaver, R.J.; Leitten, C.F. Jr.

    1962-04-17

    A boron-10 containing reactor control element wherein the boron-10 is dispersed in a matrix material is describeri. The concentration of boron-10 in the matrix varies transversely across the element from a minimum at the surface to a maximum at the center of the element, prior to exposure to neutrons. (AEC)

  6. NEUTRONIC REACTOR FUEL ELEMENT

    DOE Patents [OSTI]

    Stacy, J.T.

    1958-12-01

    A reactor fuel element having a core of molybdenum-uranium alloy jacketed in stainless steel is described. A barrier layer of tungsten, tantalum, molybdenum, columbium, or silver is interposed between the core and jacket to prevent formation of a low melting eutectic between uranium and the varlous alloy constituents of the stainless steel.

  7. Nuclear Reactors and Technology

    SciTech Connect (OSTI)

    Cason, D.L.; Hicks, S.C.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  8. NUCLEAR REACTOR COOLANT

    DOE Patents [OSTI]

    Colichman, E.L.

    1959-10-20

    The formation of new reactor coolants which suppress polymerization resulting from pyrolitic and radiation decomposition is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to about 10% of an alkall metal dispersed in the hydrocarbon.

  9. NUCLEAR REACTOR COOLANT

    DOE Patents [OSTI]

    Colichman, E.L.

    1959-10-20

    The formation of new reactor coolants which suppress polymerization resulting from pyrolytic and radiation decomposition is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to about 5% of beryllium or magnesium dispersed in the hydrocarbon.

  10. WATER BOILER REACTOR

    DOE Patents [OSTI]

    King, L.D.P.

    1960-11-22

    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.

  11. NEUTRONIC REACTOR STRUCTURE

    DOE Patents [OSTI]

    Weinberg, A.M.; Vernon, H.C.

    1961-05-30

    A neutronic reactor is described. It has a core consisting of natural uranium and heavy water and having a K-factor greater than unity which is surrounded by a reflector consisting of natural uranium and ordinary water having a Kfactor less than unity.

  12. Neutronic Reactor Structure

    DOE Patents [OSTI]

    Vernon, H. C.; Weinberg, A. M.

    1961-05-30

    The neutronic reactor is comprised of a core consisting of natural uranium and heavy water with a K-factor greater than unity. The core is surrounded by a reflector consisting of natural uranium and ordinary water with a Kfactor less than unity. (AEC)

  13. EMERGENCY SHUTDOWN FOR NUCLEAR REACTORS

    DOE Patents [OSTI]

    Paget, J.A.; Koutz, S.L.; Stone, R.S.; Stewart, H.B.

    1963-12-24

    An emergency shutdown or scram apparatus for use in a nuclear reactor that includes a neutron absorber suspended from a temperature responsive substance that is selected to fail at a preselected temperature in excess of the normal reactor operating temperature, whereby the neutron absorber is released and allowed to fall under gravity to a preselected position within the reactor core is presented. (AEC)

  14. REACTOR FUEL ELEMENTS TESTING CONTAINER

    DOE Patents [OSTI]

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  15. ALTERNATIVES TO HELIUM-3 FOR NEUTRON MULTIPLICITY DETECTORS

    SciTech Connect (OSTI)

    Ely, James H.; Siciliano, Edward R.; Swinhoe, Martyn T.

    2012-02-07

    Collaboration between the Pacific Northwest National Laboratory (PNNL) and the Los Alamos National Laboratory (LANL) is underway to evaluate neutron detection technologies that might replace the high-pressure helium (3He) tubes currently used in neutron multiplicity counter for safeguards applications. The current stockpile of 3He is diminishing and alternatives are needed for a variety of neutron detection applications including multiplicity counters. The first phase of this investigation uses a series of Monte Carlo calculations to simulate the performance of an existing neutron multiplicity counter configuration by replacing the 3He tubes in a model for that counter with candidate alternative technologies. These alternative technologies are initially placed in approximately the same configuration as the 3He tubes to establish a reference level of performance against the 3He-based system. After these reference-level results are established, the configurations of the alternative models will be further modified for performance optimization. The 3He model for these simulations is the one used by LANL to develop and benchmark the Epithermal Neutron Multiplicity Counter (ENMC) detector, as documented by H.O. Menlove, et al. in the 2004 LANL report LA-14088. The alternative technologies being evaluated are the boron-tri-fluoride-filled proportional tubes, boron-lined tubes, and lithium coated materials previously tested as possible replacements in portal monitor screening applications, as documented by R.T. Kouzes, et al. in the 2010 PNNL report PNNL-72544 and NIM A 623 (2010) 1035–1045. The models and methods used for these comparative calculations will be described and preliminary results shown

  16. Reactor vessel support system. [LMFBR

    DOE Patents [OSTI]

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  17. Strategic Need for Multi-Purpose Thermal Hydraulic Loop for Support of Advanced Reactor Technologies

    SciTech Connect (OSTI)

    James E. O'Brien; Piyush Sabharwall; Su-Jong Yoon; Gregory K. Housley

    2014-09-01

    This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

  18. Nuclear reactor construction with bottom supported reactor vessel

    DOE Patents [OSTI]

    Sharbaugh, John E.

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment

  19. Forced Two-Phase Helium Cooling Scheme for the Mu2e Transport Solenoid

    SciTech Connect (OSTI)

    Tatkowski, G.; Cheban, S.; Dhanaraj, N.; Evbota, D.; Lopes, M.; Nicol, T.; Sanders, R.; Schmitt, R.; Voirin, E.

    2015-01-01

    The Mu2e Transport Solenoid (TS) is an S-shaped magnet formed by two separate but similar magnets, TS-u and TS-d. Each magnet is quarter-toroid shaped with a centerline radius of approximately 3 m utilizing a helium cooling loop consisting of 25 to 27 horizontal-axis rings connected in series. This cooling loop configuration has been deemed adequate for cooling via forced single phase liquid helium; however it presents major challenges to forced two-phase flow such as “garden hose” pressure drop, concerns of flow separation from tube walls, difficulty of calculation, etc. Even with these disadvantages, forced two-phase flow has certain inherent advantages which make it a more attractive option than forced single phase flow. It is for this reason that the use of forced two-phase flow was studied for the TS magnets. This paper will describe the analysis using helium-specific pressure drop correlations, conservative engineering approach, helium properties calculated and updated at over fifty points, and how the results compared with those in literature. Based on the findings, the use of forced-two phase helium is determined to be feasible for steady-state cooling of the TS solenoids

  20. Helium embrittlement model and program plan for weldability of ITER materials

    SciTech Connect (OSTI)

    Louthan, M.R. Jr.; Kanne, W.R. Jr.; Tosten, M.H.; Rankin, D.T.; Cross, B.J.

    1997-02-01

    This report presents a refined model of how helium embrittles irradiated stainless steel during welding. The model was developed based on experimental observations drawn from experience at the Savannah River Site and from an extensive literature search. The model shows how helium content, stress, and temperature interact to produce embrittlement. The model takes into account defect structure, time, and gradients in stress, temperature and composition. The report also proposes an experimental program based on the refined helium embrittlement model. A parametric study of the effect of initial defect density on the resulting helium bubble distribution and weldability of tritium aged material is proposed to demonstrate the roll that defects play in embrittlement. This study should include samples charged using vastly different aging times to obtain equivalent helium contents. Additionally, studies to establish the minimal sample thickness and size are needed for extrapolation to real structural materials. The results of these studies should provide a technical basis for the use of tritium aged materials to predict the weldability of irradiated structures. Use of tritium charged and aged material would provide a cost effective approach to developing weld repair techniques for ITER components.

  1. Elucidation of fundamental properties of helium in metals by nuclear magnetic resonance techniques

    SciTech Connect (OSTI)

    Abell, G.C.

    1990-01-01

    The nuclear magnetic resonance (NMR) properties of very high density {sup 3}He in metals are discussed in the context of the corresponding properties in relatively high density bulk {sup 3}He. In particular, the effects of the {sup 3}He diffusion on the contribution of the {sup 3}He-{sup 3}He dipolar interaction to the lineshape and to the spin-lattice relaxation parameter (T{sub 1}) are described. It is shown that the temperature dependence of the lineshape and of T{sub 1} are independent sources of information about helium density and also about helium diffusivity. Moreover, T{sub 1} is shown to be a sensitive indicator of melting transitions in bulk {sup 3}He. Palladium tritide is presented as a model system for NMR studies of {sup 3}He in metals. Experimental NMR studies of this system reveal behavior analogous to what has been observed for bulk helium. Evidence for a {sup 3}He phase transition near 250 K is provided by the temperature dependence of T{sub 1}. Assuming this to be a melting transition, a density is obtained from the bulk helium EOS that is in good agreement with theory and with swelling measurements on related metal tritides. {sup 3}He NMR measurements have also provided information about the density distribution, helium diffusivity, and mean bubble size in palladium tritide. 22 refs., 8 figs.

  2. Helium release and microstructural changes in Er(D,T)2-x3Hex films).

    SciTech Connect (OSTI)

    Gelles, D. S.; Browning, James Frederick; Snow, Clark Sheldon; Banks, James Clifford; Mangan, Michael A.; Rodriguez, Mark Andrew; Brewer, Luke N.; Kotula, Paul Gabriel

    2007-12-01

    Er(D,T){sub 2-x} {sup 3}He{sub x}, erbium di-tritide, films of thicknesses 500 nm, 400 nm, 300 nm, 200 nm, and 100 nm were grown and analyzed by Transmission Electron Microscopy, X-Ray Diffraction, and Ion Beam Analysis to determine variations in film microstructure as a function of film thickness and age, due to the time-dependent build-up of {sup 3}He in the film from the radioactive decay of tritium. Several interesting features were observed: One, the amount of helium released as a function of film thickness is relatively constant. This suggests that the helium is being released only from the near surface region and that the helium is not diffusing to the surface from the bulk of the film. Two, lenticular helium bubbles are observed as a result of the radioactive decay of tritium into {sup 3}He. These bubbles grow along the [111] crystallographic direction. Three, a helium bubble free zone, or 'denuded zone' is observed near the surface. The size of this region is independent of film thickness. Four, an analysis of secondary diffraction spots in the Transmission Electron Microscopy study indicate that small erbium oxide precipitates, 5-10 nm in size, exist throughout the film. Further, all of the films had large erbium oxide inclusions, in many cases these inclusions span the depth of the film.

  3. Nuclear reactor shutdown system

    DOE Patents [OSTI]

    Bhate, Suresh K.; Cooper, Martin H.; Riffe, Delmar R.; Kinney, Calvin L.

    1981-01-01

    An inherent shutdown system for a nuclear reactor having neutron absorbing rods affixed to an armature which is held in an upper position by a magnetic flux flowing through a Curie temperature material. The Curie temperature material is fixedly positioned about the exterior of an inner duct in an annular region through which reactor coolant flows. Elongated fuel rods extending from within the core upwardly toward the Curie temperature material are preferably disposed within the annular region. Upon abnormal conditions which result in high neutron flux and coolant temperature, the Curie material loses its magnetic permeability, breaking the magnetic flux path and allowing the armature and absorber rods to drop into the core, thus shutting down the fissioning reaction. The armature and absorber rods are retrieved by lowering the housing for the electromagnet forming coils which create a magnetic flux path which includes the inner duct wall. The coil housing then is raised, resetting the armature.

  4. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Peng, Yueng-Kay M.

    1989-04-04

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  5. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Peng, Yueng-Kay M.

    1989-01-01

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  6. Neutronic reactor construction

    DOE Patents [OSTI]

    Huston, Norman E.

    1976-07-06

    1. A neutronic reactor comprising a moderator including horizontal layers formed of horizontal rows of graphite blocks, alternate layers of blocks having the rows extending in one direction, the remaining alternate layers having the rows extending transversely to the said one direction, alternate rows of blocks in one set of alternate layers having longitudinal ducts, the moderator further including slotted graphite tubes positioned in the ducts, the reactor further comprising an aluminum coolant tube positioned within the slotted tube in spaced relation thereto, bodies of thermal-neutron-fissionable material, and jackets enclosing the bodies and being formed of a corrosion-resistant material having a low neutron-capture cross section, the bodies and jackets being positioned within the coolant tube so that the jackets are spaced from the coolant tube.

  7. COMPOSITE NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Menke, J.R.

    1963-06-11

    This patent relates to a reactor having a core which comprises an inner active region and an outer active region, each region separately having a k effective less than one and a k infinity greater than one. The inner and outer regions in combination have a k effective at least equal to one and each region contributes substantially to the k effective of the reactor core. The inner region has a low moderator to fuel ratio such that the majority of fissions occurring therein are induced by neutrons having energies greater than thermal. The outer region has a high moderator to fuel ratio such that the majority of fissions occurring therein are induced by thermal neutrons. (AEC)

  8. ENGINEERING TEST REACTOR

    DOE Patents [OSTI]

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

    1958-10-21

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

  9. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, D.L.

    1987-09-04

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

  10. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, E.

    1983-08-15

    A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

  11. HOMOGENEOUS NUCLEAR REACTOR

    DOE Patents [OSTI]

    Hammond, R.P.; Busey, H.M.

    1959-02-17

    Nuclear reactors of the homogeneous liquid fuel type are discussed. The reactor is comprised of an elongated closed vessel, vertically oriented, having a critical region at the bottom, a lower chimney structure extending from the critical region vertically upwardly and surrounded by heat exchanger coils, to a baffle region above which is located an upper chimney structure containing a catalyst functioning to recombine radiolyticallydissociated moderator gages. In operation the liquid fuel circulates solely by convection from the critical region upwardly through the lower chimney and then downwardly through the heat exchanger to return to the critical region. The gases formed by radiolytic- dissociation of the moderator are carried upwardly with the circulating liquid fuel and past the baffle into the region of the upper chimney where they are recombined by the catalyst and condensed, thence returning through the heat exchanger to the critical region.

  12. AIR COOLED NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Szilard, L.

    1958-05-27

    A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

  13. LOADING MACHINE FOR REACTORS

    DOE Patents [OSTI]

    Simon, S.L.

    1959-07-01

    An apparatus is described for loading or charging slugs of fissionable material into a nuclear reactor. The apparatus of the invention is a "muzzle loading" type comprising a delivery tube or muzzle designed to be brought into alignment with any one of a plurality of fuel channels. The delivery tube is located within the pressure shell and it is also disposed within shielding barriers while the fuel cantridges or slugs are forced through the delivery tube by an externally driven flexible ram.

  14. In situ reactor

    DOE Patents [OSTI]

    Radtke, Corey William; Blackwelder, David Bradley

    2004-01-27

    An in situ reactor for use in a geological strata, is described and which includes a liner defining a centrally disposed passageway and which is placed in a borehole formed in the geological strata; and a sampling conduit is received within the passageway defined by the liner and which receives a geological specimen which is derived from the geological strata, and wherein the sampling conduit is in fluid communication with the passageway defined by the liner.

  15. BioReactor

    Energy Science and Technology Software Center (OSTI)

    2003-04-18

    BioReactor is a simulation tool kit for modeling networks of coupled chemical processes (or similar productions rules). The tool kit is implemented in C++ and has the following functionality: 1. Monte Carlo discrete event simulator 2. Solvers for ordinary differential equations 3. Genetic algorithm optimization routines for reverse engineering of models using either Monte Carlo or ODE representation )i.e., 1 or 2)

  16. REACTOR MODERATOR STRUCTURE

    DOE Patents [OSTI]

    Greenstreet, B.L.

    1963-12-31

    A system for maintaining the alignment of moderator block structures in reactors is presented. Integral restraining grids are placed between each layer of blocks in the moderator structure, at the top of the uppermost layer, and at the bottom of the lowermost layer. Slots are provided in the top and bottom surfaces of the moderator blocks so as to provide a keying action with the grids. The grids are maintained in alignment by vertical guiding members disposed about their peripheries. (AEC)

  17. A COOLED NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wigner, E.P.; Creutz, E.C.

    1960-03-15

    A nuclear reactor comprising a pair of graphite blocks separated by an air gap is described. Each of the blocks contains a plurality of channels extending from the gap through the block with a plurality of fuel elements being located in the channels. Means are provided for introducing air into the gap between the graphite blocks and for exhausting the air from the ends of the channels opposite the gap.

  18. COMPARTMENTED REACTOR FUEL ELEMENT

    DOE Patents [OSTI]

    Cain, F.M. Jr.

    1962-09-11

    A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

  19. BOILER-SUPERHEATED REACTOR

    DOE Patents [OSTI]

    Heckman, T.P.

    1961-05-01

    A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

  20. NUCLEAR REACTOR CORE DESIGN

    DOE Patents [OSTI]

    Mahlmeister, J.E.; Peck, W.S.; Haberer, W.V.; Williams, A.C.

    1960-03-22

    An improved core design for a sodium-cooled, graphitemoderated nuclear reactor is described. The improved reactor core comprises a number of blocks of moderator material, each block being in the shape of a regular prism. A number of channels, extending the length of each block, are disposed around the periphery. When several blocks are placed in contact to form the reactor core, the channels in adjacent blocks correspond with each other to form closed conduits extending the length of the core. Fuel element clusters are disposed in these closed conduits, and liquid coolant is forced through the annulus between the fuel cluster and the inner surface of the conduit. In a preferred embodiment of the invention, the moderator blocks are in the form of hexagonal prisms with longitudinal channels cut into the corners of the hexagon. The main advantage of an "edge-loaded" moderator block is that fewer thermal neutrons are absorbed by the moderator cladding, as compared with a conventional centrally loaded moderator block.

  1. Nuclear reactor sealing system

    DOE Patents [OSTI]

    McEdwards, James A.

    1983-01-01

    A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

  2. F Reactor Inspection | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    F Reactor Inspection F Reactor Inspection Addthis Description Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor last week before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has

  3. Advanced Nuclear Reactors | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Advanced Nuclear Reactors Advanced Nuclear Reactors Turbulent Flow of Coolant in an Advanced Nuclear Reactor Visualizing Coolant Flow in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor vessel where all the nuclear reactions occur to produce the heat and steam used for power generation. Nonetheless, an entire power plant consists of many other support components and key

  4. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-06-01

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full

  5. Commissioning of helium compression system for the 12 GeV refrigerator

    SciTech Connect (OSTI)

    Knudsen, Peter N.; Ganni, Venkatarao; Dixon, Kelly D.; Norton, Robert O.; Creel, Jonathan D.; Arenius, Dana M.

    2014-01-01

    The compressor system used for the Jefferson Lab (JLab) 12 GeV upgrade, also known as the CHL-2 compressor system, incorporates many design changes to the typical compressor skid design to improve the efficiency, reliability and maintainability from previous systems. These include a considerably smaller bulk oil separator design that does not use coalescing elements/media, automated control of cooling oil injection based on the helium discharge temperature, a helium after-cooler design that is designed for and promotes coalescing of residual oil and a variable speed bearing oil pump to reduce oil bypass. The CHL-2 helium compression system has five compressors configured with four pressure levels that supports the three pressure levels in the cold box. This paper will briefly review several of these improvements and discuss some of the recent commissioning results.

  6. Ab initio study of formation, migration and binding properties of helium-vacancy clusters in aluminum

    SciTech Connect (OSTI)

    Yang, Li; Zu, Xiaotao T.; Gao, Fei

    2008-08-01

    Ab initio calculations based on density functional theory have been performed to study the dissolution and migration of helium, and the stability of small helium-vacancy clusters HenVm (n, m=0 to 4) in aluminum. The results indicate that the octahedral configuration is more stable than the tetrahedral. Interstitial helium atoms are predicted to have attractive interactions and jump between two octahedral sites via an intermediate tetrahedral site with low migration energy of 0.10 eV. The binding energies of an interstitial He atom and an isolated vacancy to a HenVm cluster are also obtained from the calculated formation energies of the clusters. We find that the divacancy and tri--vacancy clusters are not stable, but He atoms can increase the stability of vacancy clusters. The interactions of He atoms with a vacancy are found to be in good agreement with the experimental results.

  7. The Advanced High-Temperature Reactor (AHTR) for Producing Hydrogen to Manufacture Liquid Fuels

    SciTech Connect (OSTI)

    Forsberg, C.W.; Peterson, P.F.; Ott, L.

    2004-10-06

    Conventional world oil production is expected to peak within a decade. Shortfalls in production of liquid fuels (gasoline, diesel, and jet fuel) from conventional oil sources are expected to be offset by increased production of fuels from heavy oils and tar sands that are primarily located in the Western Hemisphere (Canada, Venezuela, the United States, and Mexico). Simultaneously, there is a renewed interest in liquid fuels from biomass, such as alcohol; but, biomass production requires fertilizer. Massive quantities of hydrogen (H2) are required (1) to convert heavy oils and tar sands to liquid fuels and (2) to produce fertilizer for production of biomass that can be converted to liquid fuels. If these liquid fuels are to be used while simultaneously minimizing greenhouse emissions, nonfossil methods for the production of H2 are required. Nuclear energy can be used to produce H2. The most efficient methods to produce H2 from nuclear energy involve thermochemical cycles in which high-temperature heat (700 to 850 C) and water are converted to H2 and oxygen. The peak nuclear reactor fuel and coolant temperatures must be significantly higher than the chemical process temperatures to transport heat from the reactor core to an intermediate heat transfer loop and from the intermediate heat transfer loop to the chemical plant. The reactor temperatures required for H2 production are at the limits of practical engineering materials. A new high-temperature reactor concept is being developed for H2 and electricity production: the Advanced High-Temperature Reactor (AHTR). The fuel is a graphite-matrix, coated-particle fuel, the same type that is used in modular high-temperature gas-cooled reactors (MHTGRs). The coolant is a clean molten fluoride salt with a boiling point near 1400 C. The use of a liquid coolant, rather than helium, reduces peak reactor fuel and coolant temperatures 100 to 200 C relative to those of a MHTGR. Liquids are better heat transfer fluids than gases

  8. Molecular-dynamics analysis of mobile helium cluster reactions near surfaces of plasma-exposed tungsten

    SciTech Connect (OSTI)

    Hu, Lin; Maroudas, Dimitrios; Hammond, Karl D.; Wirth, Brian D.

    2015-10-28

    We report the results of a systematic atomic-scale analysis of the reactions of small mobile helium clusters (He{sub n}, 4 ≤ n ≤ 7) near low-Miller-index tungsten (W) surfaces, aiming at a fundamental understanding of the near-surface dynamics of helium-carrying species in plasma-exposed tungsten. These small mobile helium clusters are attracted to the surface and migrate to the surface by Fickian diffusion and drift due to the thermodynamic driving force for surface segregation. As the clusters migrate toward the surface, trap mutation (TM) and cluster dissociation reactions are activated at rates higher than in the bulk. TM produces W adatoms and immobile complexes of helium clusters surrounding W vacancies located within the lattice planes at a short distance from the surface. These reactions are identified and characterized in detail based on the analysis of a large number of molecular-dynamics trajectories for each such mobile cluster near W(100), W(110), and W(111) surfaces. TM is found to be the dominant cluster reaction for all cluster and surface combinations, except for the He{sub 4} and He{sub 5} clusters near W(100) where cluster partial dissociation following TM dominates. We find that there exists a critical cluster size, n = 4 near W(100) and W(111) and n = 5 near W(110), beyond which the formation of multiple W adatoms and vacancies in the TM reactions is observed. The identified cluster reactions are responsible for important structural, morphological, and compositional features in the plasma-exposed tungsten, including surface adatom populations, near-surface immobile helium-vacancy complexes, and retained helium content, which are expected to influence the amount of hydrogen re-cycling and tritium retention in fusion tokamaks.

  9. Effect of helium growth and carbon impurities on the properties of aged metal tritides

    SciTech Connect (OSTI)

    McConville, G.T.; Menke, D.A.; West, D.; Woods, C.M.

    1995-10-01

    The interaction of tritium with metals is made complex by two phenomena. The beta decay in the metal produces {sup 3}He. The helium moves to form bubbles. We shall show that the growth of the bubbles produces a two stage swelling of the metal coming first from the appearance of the helium and second from the relaxation of the lattice disorder caused by the bubble growth. The second phenomenon is the steady state ion and free radical concentration in the tritium over gas which interacts with impurities on the metal surface. We shall show that the reaction rates are much faster than for normal hydrogen cleaning. 12 refs., 7 figs., 3 tabs.

  10. SPHERICALLY SYMMETRIC NLTE MODEL ATMOSPHERES OF HOT HYDROGEN-HELIUM FIRST STARS

    SciTech Connect (OSTI)

    Kubat, Jiri

    2012-12-15

    We present results of our calculations of NLTE model stellar atmospheres for hot Population III stars composed of hydrogen and helium. We use our own computer code for the calculation of spherically symmetric NLTE model atmospheres in hydrostatic and radiative equilibrium. The model atmospheres are then used for the calculation of emergent fluxes. These fluxes serve to evaluate the flow of high-energy photons for energies higher than ionization energies of hydrogen and helium, the so-called ionizing photon fluxes. We also present the time evolution of the ionizing photon fluxes.

  11. Application of JLab 12GeV helium refrigeration system for the FRIB accelerator at MSU

    SciTech Connect (OSTI)

    Ganni, Venkatarao; Knudsen, Peter N.; Arenius, Dana M.; Casagrande, Fabio

    2014-01-01

    The planned approach to have a turnkey helium refrigeration system for the MSU-FRIB accelerator system, encompassing the design, fabrication, installation and commissioning of the 4.5-K refrigerator cold box(es), cold compression system, warm compression system, gas management, oil removal and utility/ancillary systems, was found to be cost prohibitive. Following JLab’s suggestion, MSU-FRIB accelerator management made a formal request to evaluate the applicability of the recently designed 12GeV JLab cryogenic system for this application. The following paper will outline the findings and the planned approach for the FRIB helium refrigeration system.

  12. Application of JLab 12GeV helium refrigeration system for the FRIB accelerator at MSU

    SciTech Connect (OSTI)

    Ganni, V.; Knudsen, P.; Arenius, D.; Casagrande, F.

    2014-01-29

    The planned approach to have a turnkey helium refrigeration system for the MSU-FRIB accelerator system, encompassing the design, fabrication, installation and commissioning of the 4.5-K refrigerator cold box(es), cold compression system, warm compression system, gas management, oil removal and utility/ancillary systems, was found to be cost prohibitive. Following JLab’s suggestion, MSU-FRIB accelerator management made a formal request to evaluate the applicability of the recently designed 12GeV JLab cryogenic system for this application. The following paper will outline the findings and the planned approach for the FRIB helium refrigeration system.

  13. Turning points in reactor design

    SciTech Connect (OSTI)

    Beckjord, E.S.

    1995-09-01

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

  14. Fast quench reactor and method

    DOE Patents [OSTI]

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.

    1998-05-12

    A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

  15. Massive Hanford Test Reactor Removed - Plutonium Recycle Test...

    Office of Environmental Management (EM)

    Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed from Hanford's 300 Area Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed ...

  16. Electric Power Produced from Nuclear Reactor | National Nuclear...

    National Nuclear Security Administration (NNSA)

    Electric Power Produced from Nuclear Reactor Electric Power Produced from Nuclear Reactor Arco, ID The Experimental Breeder Reactor No. 1 located at the National Reactor Testing ...

  17. A Preliminary Investigation of Rapid Depressurization Phenomena Following a Sudden DLOFC in a VHTR

    SciTech Connect (OSTI)

    Richard C. Martineau; Ray A. Berry; Dana A. Knoll

    2009-03-01

    Air ingress has been identified as a potential threat for Very High Temperature gas-cooled Reactors (VHTR). Reactor components constructed of graphite will, at high temperatures, produce exothermic reactions in the presence of oxygen. The danger lies in the possibility of fuel element damage and core structural failure. Previous investigations of air ingress mechanisms have focused on thermal and molecular diffusion, density-driven stratified flow, and natural convection. Here, we investigate the possibility of a rapid ingress of air due to a Taylor wave expansion after a hypothetical sudden loss of coolant accident (LOCA) scenario in a VHTR. Our analysis starts with a one-dimensional shock tube simulation to simply illustrate the development of a Taylor wave with resulting reentrant flow. Then, a simulation is performed of an idealized two-dimensional axisymmetric representation of the lower plenum of General Atomics GT-MHR subjected to a hypothetical catastrophic break of the hot duct. Analysis shows the potential for significant and rapid air ingress into the reactor vessel in the case of a large break in the cooling system.

  18. Progress Update: Reactor Disassembly Grouting

    SciTech Connect (OSTI)

    Cody, Tom

    2010-01-01

    Grouting the P&R reactors in order to remove these basins as an environmental threat. This will end the Cold War legacy and end the environmental footprint.

  19. Reactor Materials Newsletter- Issue 1

    Energy.gov [DOE]

    The Reactor Materials (RM) newsletter includes information about key nuclear materials programs, results from ongoing projects across the Office of Nuclear Energy, and other relevant information.

  20. Daya Bay Reactor Neutrino Experiment

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Ao Nuclear Power Plant reactors. The experiment is being built by blasting three kilometers of tunnel through the granite rock under the mountains where the power plants are...

  1. Progress Update: Reactor Disassembly Grouting

    ScienceCinema (OSTI)

    Cody, Tom

    2012-06-14

    Grouting the P&R reactors in order to remove these basins as an environmental threat. This will end the Cold War legacy and end the environmental footprint.

  2. Neutrino oscillation studies with reactors

    DOE PAGES-Beta [OSTI]

    Vogel, P.; Wen, L.J.; Zhang, C.

    2015-04-27

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavours are quantum mechanical mixtures. Over the past several decades, reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle θ13. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  3. Effects of helium implantation on the tensile properties and microstructure of Ni??P?? metallic glass nanostructures

    SciTech Connect (OSTI)

    Liontas, Rachel; Gu, X. Wendy; Fu, Engang; Wang, Yongqiang; Li, Nan; Mara, Nathan; Greer, Julia R.

    2014-09-10

    We report fabrication and nanomechanical tension experiments on as-fabricated and helium-implanted ~130 nm diameter Ni??P?? metallic glass nano-cylinders. The nano-cylinders were fabricated by a templated electroplating process and implanted with He? at energies of 50, 100, 150, and 200 keV to create a uniform helium concentration of ~3 at. % throughout the nano-cylinders. Transmission electron microscopy (TEM) imaging and through-focus analysis reveal that the specimens contained ~2 nm helium bubbles distributed uniformly throughout the nano-cylinder volume. In-situ tensile experiments indicate that helium-implanted specimens exhibit enhanced ductility as evidenced by a 2-fold increase in plastic strain over as-fabricated specimens, with no sacrifice in yield and ultimate tensile strengths. This improvement in mechanical properties suggests that metallic glasses may actually exhibit a favorable response to high levels of helium implantation.

  4. Effects of helium implantation on the tensile properties and microstructure of Ni₇₃P₂₇ metallic glass nanostructures

    DOE PAGES-Beta [OSTI]

    Liontas, Rachel; Gu, X. Wendy; Fu, Engang; Wang, Yongqiang; Li, Nan; Mara, Nathan; Greer, Julia R.

    2014-09-10

    We report fabrication and nanomechanical tension experiments on as-fabricated and helium-implanted ~130 nm diameter Ni₇₃P₂₇ metallic glass nano-cylinders. The nano-cylinders were fabricated by a templated electroplating process and implanted with He⁺ at energies of 50, 100, 150, and 200 keV to create a uniform helium concentration of ~3 at. % throughout the nano-cylinders. Transmission electron microscopy (TEM) imaging and through-focus analysis reveal that the specimens contained ~2 nm helium bubbles distributed uniformly throughout the nano-cylinder volume. In-situ tensile experiments indicate that helium-implanted specimens exhibit enhanced ductility as evidenced by a 2-fold increase in plastic strain over as-fabricated specimens, with nomore » sacrifice in yield and ultimate tensile strengths. This improvement in mechanical properties suggests that metallic glasses may actually exhibit a favorable response to high levels of helium implantation.« less

  5. Nuclear reactor control apparatus

    DOE Patents [OSTI]

    Sridhar, Bettadapur N.

    1983-11-01

    Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additional magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

  6. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.

    1993-12-14

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase is described. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figures.

  7. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

    1995-01-01

    A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves.

  8. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN)

    1993-01-01

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves.

  9. FOOD IRRADIATION REACTOR

    DOE Patents [OSTI]

    Leyse, C.F.; Putnam, G.E.

    1961-05-01

    An irradiation apparatus is described. It comprises a pressure vessel, a neutronic reactor active portion having a substantially greater height than diameter in the pressure vessel, an annular tank surrounding and spaced from the pressure vessel containing an aqueous indium/sup 1//sup 1//sup 5/ sulfate solution of approximately 600 grams per liter concentration, means for circulating separate coolants through the active portion and the space between the annular tank and the pressure vessel, radiator means adapted to receive the materials to be irradiated, and means for flowing the indium/sup 1//sup 1//sup 5/ sulfate solution through the radiator means.

  10. NEUTRONIC REACTOR CONSTRUCTION

    DOE Patents [OSTI]

    Vernon, H.C.; Goett, J.J.

    1958-09-01

    A cover device is described for the fuel element receiving tube of a neutronic reactor of the heterogeneous, water cooled type wherein said tubes are arranged in a moderator with their longitudinal axes vertical. The cover is provided with means to support a rod-type fuel element from the bottom thereof and means to lock the cover in place, the latter being adapted for remote operation. This cover device is easily removable and seals the opening in the upper end of the fuel tube against leakage of coolant.

  11. Reactor coolant pump flywheel

    SciTech Connect (OSTI)

    Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

    2013-11-26

    A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

  12. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

    1996-01-01

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves.

  13. PINCHED PLASMA REACTOR

    DOE Patents [OSTI]

    Phillips, J.A.; Suydam, R.; Tuck, J.L.

    1961-07-01

    BS>A plasma confining and heating reactor is described which has the form of a torus with a B/sub 2/ producing winding on the outside of the torus and a helical winding of insulated overlapping tunns on the inside of the torus. The inner helical winding performs the double function of shielding the plasma from the vitreous container and generating a second B/sub z/ field in the opposite direction to the first B/sub z/ field after the pinch is established.

  14. REACTOR COOLANT TUBE SEAL

    DOE Patents [OSTI]

    Morris, W.J.

    1958-12-01

    A plle-flattenlng control element and a fluid seal therefore to permit movement of the element into a liquld contnining region of a neutronlc reactor are described. The device consists of flattened, thin-walled aluminum tubing contalnlng a uniform mixture of thermal neutron absorbing material, and a number of soft rubber closures for the process tubes, having silts capable of passing the flattened elements therethrough, but effectively sealing the process tubes against fluld leaknge by compression of the rubber. The flattened tubing is sufficiently flexible to enable it to conform to the configuratlon of the annular spacing surrounding the fuel elements ln the process tubes.

  15. High flux reactor

    DOE Patents [OSTI]

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

    1988-01-01

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

  16. Fast Reactor Technology Preservation

    SciTech Connect (OSTI)

    Wootan, David W.; Omberg, Ronald P.

    2008-01-11

    There is renewed worldwide interest in developing and implementing a new generation of advanced fast reactors. International cooperative efforts are underway such as the Global Nuclear Energy Partnership (GNEP). Advanced computer modeling and simulation efforts are a key part of these programs. A recognized and validated set of Benchmark Cases are an essential component of such modeling efforts. Testing documentation developed during the operation of the Fast Flux Test Facility (FFTF) provide the information necessary to develop a very useful set of Benchmark Cases.

  17. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, James E.; Meuschke, Robert E.

    1995-01-01

    A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

  18. POWER BREEDER REACTOR

    DOE Patents [OSTI]

    Monson, H.O.

    1960-11-22

    An arrangement is offered for preventing or minimizing the contraction due to temperature rise, of a reactor core comprising vertical fuel rods in sodium. Temperature rise of the fuel rods would normally make them move closer together by inward bowing, with a resultant undesired increase in reactivity. According to the present invention, assemblies of the fuel rods are laterally restrained at the lower ends of their lower blanket sections and just above the middle of the fuel sections proper of the rods, and thus the fuel sections move apart, rather than together, with increase in temperature.

  19. NUCLEAR REACTOR FUEL ELEMENT

    DOE Patents [OSTI]

    Anderson, W.F.; Tellefson, D.R.; Shimazaki, T.T.

    1962-04-10

    A plate type fuel element which is particularly useful for organic cooled reactors is described. Generally, the fuel element comprises a plurality of fissionable fuel bearing plates held in spaced relationship by a frame in which the plates are slidably mounted in grooves. Clearance is provided in the grooves to allow the plates to expand laterally. The plates may be rigidly interconnected but are floatingly supported at their ends within the frame to allow for longi-tudinal expansion. Thus, this fuel element is able to withstand large temperature differentials without great structural stresses. (AEC)

  20. FAST NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Snell, A.H.

    1957-12-01

    This patent relates to a reactor and process for carrying out a controlled fast neutron chain reaction. A cubical reactive mass, weighing at least 920 metric tons, of uranium metal containing predominantly U/sup 238/ and having a U/sup 235/ content of at least 7.63% is assembled and the maximum neutron reproduction ratio is limited to not substantially over 1.01 by insertion and removal of a varying amount of boron, the reactive mass being substantially freed of moderator.

  1. Hydroprocessing reactors and methods

    SciTech Connect (OSTI)

    Bridge, A.G.; Strangeland, B.E.

    1986-10-07

    This patent describes a downflow hydrocarbon hydroprocessing reactor containing vertically spaced catalyst beds. The improvement described here comprises: means for increasing the contacting time between hydrocarbon vapor and the catalyst including; a liquid/vapor separator placed between two of the catalyst beds. The separator comprises a bypass means so arranged that liquid separated by the separator will enter the bypass means and not contact the catalyst bed or beds below the separator while the vapor separated by the separator will pass through the catalyst bed or beds below the separator.

  2. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.; Marasco, J.A.

    1996-02-27

    A fluidized bed reactor system is described which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves. 3 figs.

  3. Nuclear reactor fuel element

    DOE Patents [OSTI]

    Johnson, Carl E.; Crouthamel, Carl E.

    1980-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.

  4. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.; Marasco, J.A.

    1995-04-25

    A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figs.

  5. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, J.E.; Meuschke, R.E.

    1995-05-02

    A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

  6. Emergency Decay Heat Removal in a GEN-IV Gas-Cooled Fast Reactor

    SciTech Connect (OSTI)

    Cheng, Lap Y.; Ludewig, Hans; Jo, Jae [Brookhaven National Laboratory, P.O. Box 5000, Upton, NY 11973-5000 (United States)

    2006-07-01

    A series of transient analyses using the system code RELAP5-3d has been performed to confirm the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400 MWt GEN-IV gas-cooled fast reactor. The accident sequence of interest is a station blackout simultaneous with a small break (10 sq.inch/0.645 m{sup 2}) in the reactor vessel. The analyses cover the three phases of decay heat removal in a depressurization accident: (1) forced flow cooling by the power conversion unit (PCU) coast down, (2) active forced flow cooling by a battery powered blower, and (3) passive cooling by natural circulation. The blower is part of an emergency cooling system (ECS) that by design is to sustain passive decay heat removal via natural circulation cooling 24 hours after shutdown. The RELAP5 model includes the helium-cooled reactor, the ECS (primary and secondary side), the PCU with all the rotating machinery (turbine and compressors) and the heat transfer components (recuperator, pre-cooler and inter-cooler), and the guard containment that surrounds the reactor and the PCU. The transient analysis has demonstrated the effectiveness of passive decay heat removal by natural circulation cooling when the guard containment pressure is maintained at or above 800 kPa. (authors)

  7. A Conceptual Multi-Megawatt System Based on a Tungsten CERMET Reactor

    SciTech Connect (OSTI)

    Jonathan A. Webb; Brian Gross

    2011-02-01

    Abstract. A conceptual reactor system to support Multi-Megawatt Nuclear Electric Propulsion is investigated within this paper. The reactor system consists of a helium cooled Tungsten-UN fission core, surrounded by a beryllium neutron reflector and 13 B4C control drums coupled to a high temperature Brayton power conversion system. Excess heat is rejected via carbon reinforced heat pipe radiators and the gamma and neutron flux is attenuated via segmented shielding consisting of lithium hydride and tungsten layers. Turbine inlet temperatures ranging from 1300 K to 1500 K are investigated for their effects on specific powers and net electrical outputs ranging from 1 MW to 100 MW. The reactor system is estimated to have a mass, which ranges from 15 Mt at 1 MWe and a turbine inlet temperature of 1500 K to 1200 Mt at 100 MWe and a turbine temperature of 1300 K. The reactor systems specific mass ranges from 32 kg/kWe at a turbine inlet temperature of 1300 K and a power of 1 MWe to 9.5 kg/kW at a turbine temperature of 1500 K and a power of 100 MWe.

  8. Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

    Office of Energy Efficiency and Renewable Energy (EERE)

    Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary November 2014

  9. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect (OSTI)

    Not Available

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

  10. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect (OSTI)

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of inherent safety concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and thermophysical

  11. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    SciTech Connect (OSTI)

    Peterson, Per; Greenspan, Ehud

    2015-02-09

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X

  12. Reactor refueling machine simulator

    SciTech Connect (OSTI)

    Rohosky, T.L.; Swidwa, K.J.

    1987-10-13

    This patent describes in combination: a nuclear reactor; a refueling machine having a bridge, trolley and hoist each driven by a separate motor having feedback means for generating a feedback signal indicative of movement thereof. The motors are operable to position the refueling machine over the nuclear reactor for refueling the same. The refueling machine also has a removable control console including means for selectively generating separate motor signals for operating the bridge, trolley and hoist motors and for processing the feedback signals to generate an indication of the positions thereof, separate output leads connecting each of the motor signals to the respective refueling machine motor, and separate input leads for connecting each of the feedback means to the console; and a portable simulator unit comprising: a single simulator motor; a single simulator feedback signal generator connected to the simulator motor for generating a simulator feedback signal in response to operation of the simulator motor; means for selectively connecting the output leads of the console to the simulator unit in place of the refueling machine motors, and for connecting the console input leads to the simulator unit in place of the refueling machine motor feedback means; and means for driving the single simulator motor in response to any of the bridge, trolley or hoist motor signals generated by the console and means for applying the simulator feedback signal to the console input lead associated with the motor signal being generated by the control console.

  13. NEUTRONIC REACTOR CORE

    DOE Patents [OSTI]

    Thomson, W.B.; Corbin, A. Jr.

    1961-07-18

    An improved core for a gas-cooled power reactor which admits gas coolant at high temperatures while affording strong integral supporting structure and efficient moderation of neutrons is described. The multiplicities of fuel elements constituting the critical amassment of fissionable material are supported and confined by a matrix of metallic structure which is interspersed therebetween. Thermal insulation is interposed between substantially all of the metallic matrix and the fuel elements; the insulation then defines the principal conduit system for conducting the coolant gas in heat-transfer relationship with the fuel elements. The metallic matrix itseif comprises a system of ducts through which an externally-cooled hydrogeneous liquid, such as water, is circulated to serve as the principal neutron moderant for the core and conjointly as the principal coolant for the insulated metallic structure. In this way, use of substantially neutron transparent metals, such as aluminum, becomes possible for the supporting structure, despite the high temperatures of the proximate gas. The Aircraft Nuclear Propulsion program's "R-1" reactor design is a preferred embodiment.

  14. Multimegawatt Space Reactor Safety

    SciTech Connect (OSTI)

    Stanley, M.L. )

    1989-01-01

    The Multimegawatt (MMW) Space Reactor Project supports the Strategic Defense Initiative Office requirement to provide reliable, safe, cost-effective, electrical power in the MMW range. Specifically, power may be used for neutral particle beams, free electron lasers, electromagnetic launchers, and orbital transfer vehicles. This power plant technology may also apply to the electrical power required for other uses such as deep-space probes and planetary exploration. The Multimegawatt Space Reactor Project, the Thermionic Fuel Element Verification Program, and Centaurus Program all support the Multimegawatt Space Nuclear Power Program and form an important part of the US Department of Energy's (DOE's) space and defense power systems activities. A major objective of the MMW project is the development of a reference flight system design that provides the desired levels of public safety, health protection, and special nuclear material (SNM) protection when used during its designated missions. The safety requirements for the MMW project are a hierarchy of requirements that consist of safety requirements/regulations, a safety policy, general safety criteria, safety technical specifications, safety design specifications, and the system design. This paper describes the strategy and philosophy behind the development of the safety requirements imposed upon the MMW concept developers. The safety organization, safety policy, generic safety issues, general safety criteria, and the safety technical specifications are discussed.

  15. Ordinary SQUID interferometers and superfluid helium matter wave interferometers: The role of quantum fluctuations

    SciTech Connect (OSTI)

    Golovashkin, A. I.; Zherikhina, L. N. Tskhovrebov, A. M.; Izmailov, G. N.; Ozolin, V. V.

    2010-08-15

    When comparing the operation of a superfluid helium matter wave quantum interferometer (He SQUID) with that of an ordinary direct-current quantum interferometer (dc SQUID), we estimate their resolution limitation that correspond to quantum fluctuations. An alternative mode of operation of the interferometer as a unified macroquantum system is considered.

  16. Simulation of streamers propagating along helium jets in ambient air: Polarity-induced effects

    SciTech Connect (OSTI)

    Naidis, G. V.

    2011-04-04

    Results of modeling of streamer propagation along helium jets for both positive and negative polarities of applied voltage are presented. Obtained patterns of streamer dynamics and structure in these two cases are similar to those observed in experiments with plasma jets.

  17. RECONCILING THE GALACTIC BULGE TURNOFF AGE DISCREPANCY WITH ENHANCED HELIUM ENRICHMENT

    SciTech Connect (OSTI)

    Nataf, David M.; Gould, Andrew P.

    2012-06-01

    We show that the factor {approx}2 discrepancy between spectroscopic and photometric age determinations of the Galactic bulge main-sequence turnoff can be naturally explained by positing an elevated helium enrichment for the bulge relative to that assumed by standard isochrones. Helium enhancement relative to standard isochrones is confirmed at the 2.3{sigma} level. We obtain an upper bound on the helium enrichment for the metal-rich ([Fe/H] Almost-Equal-To +0.30) stars of {Delta}Y Almost-Equal-To +0.11 relative to canonical expectations, given the requirement that the spectroscopic and photometric ages be consistent and the limiting condition of instantaneous star formation. We discuss phenomenological evidence that the bulge may have had a chemical evolution that is distinct from the solar neighborhood in this manner, and we make several testable predictions. Should this emerging picture of the bulge as helium-enhanced hold, it will require the development of new isochrones, new model atmospheres, and modified analysis and cosmological interpretation of the integrated light of other bulges and elliptical galaxies.

  18. A comparison of hydrogen vs. helium glow discharge effects on fusion device first-wall conditioning

    SciTech Connect (OSTI)

    Dylla, H.F.

    1989-09-01

    Hydrogen- and deuterium-fueled glow discharges are used for the initial conditioning of magnetic fusion device vacuum vessels following evacuation from atmospheric pressure. Hydrogenic glow discharge conditioning (GDC) significantly reduces the near-surface concentration of simple adsorbates, such as H/sub 2/O, CO, and CH/sub 4/, and lowers ion-induced desorption coefficients by typically three orders of magnitude. The time evolution of the residual gas production observed during hydrogen-glow discharge conditioning of the carbon first-wall structure of the TFTR device is similar to the time evolution observed during hydrogen GDC of the initial first-wall configuration in TFTR, which was primarily stainless steel. Recently, helium GDC has been investigated for several wall-conditioning tasks on a number of tokamaks including TFTR. Helium GDC shows negligible impurity removal with stainless steel walls. For impurity conditioning with carbon walls, helium GDC shows significant desorption of H/sub 2/O, CO, and CO/sub 2/; however, the total desorption yield is limited to the monolayer range. In addition, helium GDC can be used to displace hydrogen isotopes from the near-surface region of carbon first-walls in order to lower hydrogenic retention and recycling. 38 refs., 6 figs.

  19. Non-local thermodynamic equilibrium effects on isentropic coefficient in argon and helium thermal plasmas

    SciTech Connect (OSTI)

    Sharma, Rohit; Singh, Kuldip

    2014-03-15

    In the present work, two cases of thermal plasma have been considered; the ground state plasma in which all the atoms and ions are assumed to be in the ground state and the excited state plasma in which atoms and ions are distributed over various possible excited states. The variation of Z?, frozen isentropic coefficient and the isentropic coefficient with degree of ionization and non-equilibrium parameter ?(= T{sub e}/T{sub h}) has been investigated for the ground and excited state helium and argon plasmas at pressures 1?atm, 10?atm, and 100?atm in the temperature range from 6000?K to 60?000?K. For a given value of non-equilibrium parameter, the relationship of Z? with degree of ionization does not show any dependence on electronically excited states in helium plasma whereas in case of argon plasma this dependence is not appreciable till degree of ionization approaches 2. The minima of frozen isentropic coefficient shifts toward lower temperature with increase of non-equilibrium parameter for both the helium and argon plasmas. The lowering of non-equilibrium parameter decreases the frozen isentropic coefficient more emphatically in helium plasma at high pressures in comparison to argon plasma. The increase of pressure slightly reduces the ionization range over which isentropic coefficient almost remains constant and it does not affect appreciably the dependence of isentropic coefficient on non-equilibrium parameter.

  20. Helium bubble formation in ultrafine and nanocrystalline tungsten under different extreme conditions

    SciTech Connect (OSTI)

    El-atwani, O.; Hattar, Khalid Mikhiel; Hinks, J. A.; Greaves, G.; Harilal, S. S.; Hassanein, A.

    2014-12-25

    We investigated the effects of helium ion irradiation energy and sample temperature on the performance of grain boundaries as helium sinks in ultrafine grained and nanocrystalline tungsten. Irradiations were performed at displacement and non-displacement energies and at temperatures above and below that required for vacancy migration. Microstructural investigations were performed using Transmission Electron Microscopy (TEM) combined with either in-situ or ex-situ ion irradiation. Under helium irradiation at an energy which does not cause atomic displacements in tungsten (70 eV), regardless of temperature and thus vacancy migration conditions, bubbles were uniformly distributed with no preferential bubble formation on grain boundaries. Moreover, at energies that can cause displacements, bubbles were observed to be preferentially formed on the grain boundaries only at high temperatures where vacancy migration occurs. Under these conditions, the decoration of grain boundaries with large facetted bubbles occurred on nanocrystalline grains with dimensions less than 60 nm. Finally, we discuss the importance of vacancy supply and the formation and migration of radiation-induced defects on the performance of grain boundaries as helium sinks and the resulting irradiation tolerance of ultrafine grained and nanocrystalline tungsten to bubble formation.

  1. Helium bubble formation in ultrafine and nanocrystalline tungsten under different extreme conditions

    DOE PAGES-Beta [OSTI]

    El-atwani, O.; Hattar, Khalid Mikhiel; Hinks, J. A.; Greaves, G.; Harilal, S. S.; Hassanein, A.

    2014-12-25

    We investigated the effects of helium ion irradiation energy and sample temperature on the performance of grain boundaries as helium sinks in ultrafine grained and nanocrystalline tungsten. Irradiations were performed at displacement and non-displacement energies and at temperatures above and below that required for vacancy migration. Microstructural investigations were performed using Transmission Electron Microscopy (TEM) combined with either in-situ or ex-situ ion irradiation. Under helium irradiation at an energy which does not cause atomic displacements in tungsten (70 eV), regardless of temperature and thus vacancy migration conditions, bubbles were uniformly distributed with no preferential bubble formation on grain boundaries. Moreover,more » at energies that can cause displacements, bubbles were observed to be preferentially formed on the grain boundaries only at high temperatures where vacancy migration occurs. Under these conditions, the decoration of grain boundaries with large facetted bubbles occurred on nanocrystalline grains with dimensions less than 60 nm. Finally, we discuss the importance of vacancy supply and the formation and migration of radiation-induced defects on the performance of grain boundaries as helium sinks and the resulting irradiation tolerance of ultrafine grained and nanocrystalline tungsten to bubble formation.« less

  2. Statistical Properties of Inter-Series Mixing in Helium: From Integrability to Chaos

    SciTech Connect (OSTI)

    Pu''ttner, R.; Gremaud, B.; Delande, D.; Domke, M.; Martins, M.; Schlachter, A. S.; Kaindl, G.

    2001-04-23

    The photoionization spectrum of helium shows considerable complexity close to the double-ionization threshold. By analyzing the results from both our recent experiments and ab initio three- and one-dimensional calculations, we show that the statistical properties of the spacings between neighboring energy levels clearly display a transition towards quantum chaos.

  3. The First Reactor | Department of Energy

    Office of Environmental Management (EM)

    The First Reactor The First Reactor Chicago Pile-1 (CP-1) was the world's first nuclear reactor. CP-1 was built on a rackets court, under the abandoned west stands of the original ...

  4. Reactor production of Thoruim-229

    DOE PAGES-Beta [OSTI]

    Boll, Rose Ann; Murphy, Karen E.; Denton, David L.; Tamara J. Haverlock; Garland, Marc A.; Mirzadeh, Saed; Hogle, Susan; Owens, Allison

    2016-05-03

    Limited availability of 229Th for clinical applications of 213Bi necessitates investigation of alternative production routes. In reactor production, 229Th is produced from neutron transmutation of 226Ra, 228Ra, 227Ac and 228Th. Here, we evaluate irradiations of 226Ra, 228Ra, and 227Ac targets at the ORNL High Flux Isotope Reactor.

  5. The International Reactor Dosimetry File.

    Energy Science and Technology Software Center (OSTI)

    1994-01-19

    Version 01 The International Reactor Dosimetry File (IRDF-90) contains recommended neutron cross-section data to be used for reactor neutron dosimetry by foil activation. It also contains selected recommended values for radiation damage cross-sections and benchmark neutron spectra. This library supersedes all earlier versions of IRDF.

  6. NOVEL CRYOGENIC ENGINEERING SOLUTIONS FOR THE NEW AUSTRALIAN RESEARCH REACTOR OPAL

    SciTech Connect (OSTI)

    Olsen, S. R.; Kennedy, S. J.; Kim, S.; Schulz, J. C.; Thiering, R.; Gilbert, E. P.; Lu, W.; James, M.; Robinson, R. A.

    2008-03-16

    In August 2006 the new 20MW low enriched uranium research reactor OPAL went critical. The reactor has 3 main functions, radio pharmaceutical production, silicon irradiation and as a neutron source. Commissioning on 7 neutron scattering instruments began in December 2006. Three of these instruments (Small Angle Neutron Scattering, Reflectometer and Time-of-flight Spectrometer) utilize cold neutrons.The OPAL Cold Neutron Source, located inside the reactor, is a 20L liquid deuterium moderated source operating at 20K, 330kPa with a nominal refrigeration capacity of 5 kW and a peak flux at 4.2meV (equivalent to a wavelength of 0.4nm). The Thermosiphon and Moderator Chamber are cooled by helium gas delivered at 19.8K using the Brayton cycle. The helium is compressed by two 250kW compressors (one with a variable frequency drive to lower power consumption).A 5 Tesla BSCCO (2223) horizontal field HTS magnet will be delivered in the 2{sup nd} half of 2007 for use on all the cold neutron instruments. The magnet is cooled by a pulse tube cryocooler operating at 20K. The magnet design allows for the neutron beam to pass both axially and transverse to the field. Samples will be mounted in a 4K to 800K Gifford-McMahon (GM) cryofurnace, with the ability to apply a variable electric field in-situ. The magnet is mounted onto a tilt stage. The sample can thus be studied under a wide variety of conditions.A cryogen free 7.4 Tesla Nb-Ti vertical field LTS magnet, commissioned in 2005 will be used on neutron diffraction experiments. It is cooled by a standard GM cryocooler operating at 4.2K. The sample is mounted in a 2{sup nd} GM cryocooler (4K-300K) and a variable electric field can be applied.

  7. Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen Production Plant

    SciTech Connect (OSTI)

    E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

    2008-08-01

    A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540°C and 900°C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating-current, AC, to direct-current, DC, conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

  8. Repair welding of fusion reactor components

    SciTech Connect (OSTI)

    Chin, B.A.

    1993-05-15

    Experiments have shown that irradiated Type 316 stainless steel is susceptible to heat-affected-zone (HAZ) cracking upon cooling when welded using the gas tungsten arc (GTA) process under lateral constraint. The cracking has been hypothesized to be caused by stress-assisted helium bubble growth and rupture at grain boundaries. This study utilized an experimental welding setup which enabled different compressive stresses to be applied to the plates during welding. Autogenous GTA welds were produced in Type 316 stainless steel doped with 256 appm helium. The application of a compressive stress, 55 MPa, during welding suppressed the previously observed catastrophic cracking. Detailed examinations conducted after welding showed a dramatic change in helium bubble morphology. Grain boundary bubble growth along directions parallel to the weld was suppressed. Results suggest that stress-modified welding techniques may be used to suppress or eliminate helium-induced cracking during joining of irradiated materials.

  9. Solvent refined coal reactor quench system

    DOE Patents [OSTI]

    Thorogood, Robert M.

    1983-01-01

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

  10. Solvent refined coal reactor quench system

    DOE Patents [OSTI]

    Thorogood, R.M.

    1983-11-08

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

  11. Reactor Engineering Design | netl.doe.gov

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Reactor Engineering Design The Reactor Engineering Design Key Technology will focus on control of chemical reactions with unprecedented precision in increasingly modular and efficient reactors, allowing for smaller reactors and streamlined processes that will convert coal into valuable products at low cost and with high energy efficiency. Here, the specific emphasis will be reactors enabling conversion of coal-biomass to liquid fuels, Novel reactors, advanced manufacturing, etc. will be

  12. RADIATION FACILITY FOR NUCLEAR REACTORS

    DOE Patents [OSTI]

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

    1961-12-12

    A radiation facility is designed for irradiating samples in close proximity to the core of a nuclear reactor. The facility comprises essentially a tubular member extending through the biological shield of the reactor and containing a manipulatable rod having the sample carrier at its inner end, the carrier being longitudinally movable from a position in close proximity to the reactor core to a position between the inner and outer faces of the shield. Shield plugs are provided within the tubular member to prevent direct radiation from the core emanating therethrough. In this device, samples may be inserted or removed during normal operation of the reactor without exposing personnel to direct radiation from the reactor core. A storage chamber is also provided within the radiation facility to contain an irradiated sample during the period of time required to reduce the radioactivity enough to permit removal of the sample for external handling. (AEC)

  13. Fast reactors and nuclear nonproliferation

    SciTech Connect (OSTI)

    Avrorin, E.N.; Rachkov, V.I.; Chebeskov, A.N.

    2013-07-01

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

  14. Nuclear Reactor Kinetics and Control.

    Energy Science and Technology Software Center (OSTI)

    2009-07-27

    Version 00 Dr. J.D. Lewins has now released the following legacy book for free distribution: Nuclear Reactor Kinetics and Control, Pergamon Press, London, 275 pages, 1978. 1. Introductory Review 2. Neutron and Precursor Equations 3. Elementary Solutions of the Kinetics Equations at Low Power 4. Linear Reactor Process Dynamics with Feedback 5. Power Reactor Control Systems 6. Fluctuations and Reactor Noise 7. Safety and Reliability 8. Non Linear Systems; Stability and Control 9. Analogue Computingmore » Addendum: Jay Basken and Jeffery D. Lewins: Power Series Solution of the Reactor Kinetics Equations, Nuclear Science and Engineering: 122, 407-436 (1996) (authorized for distribution with the book: courtesy of the American Nuclear Society)« less

  15. Reactor-Produced Medical Radionuclides

    SciTech Connect (OSTI)

    Mirzadeh, Saed; Mausner, Leonard; Garland, Marc A

    2011-01-01

    The therapeutic use of radionuclides in nuclear medicine, oncology and cardiology is the most rapidly growing use of medical radionuclides. Since most therapeutic radionuclides are neutron rich and decay by beta emission, they are reactor-produced. This chapter deals mainly with production approaches with neutrons. Neutron interactions with matter, neutron transmission and activation rates, and neutron spectra of nuclear reactors are discussed in some detail. Further, a short discussion of the neutron-energy dependence of cross sections, reaction rates in thermal reactors, cross section measurements and flux monitoring, and general equations governing the reactor production of radionuclides are presented. Finally, the chapter is concluded by providing a number of examples encompassing the various possible reaction routes for production of a number of medical radionuclides in a reactor.

  16. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    SciTech Connect (OSTI)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  17. MOOSE simulating nuclear reactor CRUD buildup

    SciTech Connect (OSTI)

    2014-02-06

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  18. Electricity Generating Portfolios with Small Modular Reactors...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Electricity Generating Portfolios with Small Modular Reactors Electricity Generating Portfolios with Small Modular Reactors This paper provides a method for estimating the ...

  19. Manhattan Project: F Reactor Plutonium Production Complex

    Office of Scientific and Technical Information (OSTI)

    F REACTOR PLUTONIUM PRODUCTION COMPLEX Hanford Engineer Works, 1945 Resources > Photo Gallery Plutonium production area, Hanford, ca. 1945 The F Reactor plutonium production ...

  20. naval reactors | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    on Energy and Water Development, visited the Naval Reactors Facility (NRF) at the... ... propulsion plants use a pressurized-water reactor design that has two basic systems: ...

  1. Light Water Reactor Sustainability Nondestructive Evaluation...

    Energy.gov (indexed) [DOE]

    US Department of Energy Office of Nuclear Energy's Light Water Reactor Sustainability ... A multitude of concrete-based structures are typically part of a light water reactor (LWR) ...

  2. Light Water Reactor Sustainability Program - Integrated Program...

    Office of Environmental Management (EM)

    Program - Integrated Program Plan Light Water Reactor Sustainability Program - Integrated Program Plan The Light Water Reactor Sustainability (LWRS) Program is a research and ...

  3. Reactor Materials Newsletter, Issue 2, May 2016

    Energy.gov [DOE]

    Reactor Materials Newsletter - Issue 2 The Reactor Materials (RM) newsletter includes information about key nuclear materials programs and results from ongoing projects across the Office of Nuclear Energy.

  4. Nuclear power reactor instrumentation systems handbook. Volume...

    Office of Scientific and Technical Information (OSTI)

    Nuclear power reactor instrumentation systems handbook. Volume 1 Citation Details In-Document Search Title: Nuclear power reactor instrumentation systems handbook. Volume 1 You ...

  5. MOOSE simulating nuclear reactor CRUD buildup

    ScienceCinema (OSTI)

    None

    2014-07-21

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  6. Advanced Reactor Technology Documents | Department of Energy

    Energy.gov (indexed) [DOE]

    Reactor Concepts Technical Review Panel Report This report documents the establishment of a technical review process and the findings of the Advanced Reactor Concepts (ARC) ...

  7. Reactor Engineering Design | netl.doe.gov

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Reactor Engineering Design The Reactor Engineering Design Key Technology will focus on control of chemical reactions with unprecedented precision in increasingly modular and ...

  8. Advanced Reactor Thermal Hydraulic Modeling | Argonne Leadership...

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Reactor Thermal Hydraulic Modeling PI Name: Paul Fischer PI Email: fischer@mcs.anl.gov ... Advanced simulation is viewed as critical in bringing fast reactor technology to fruition ...

  9. Nuclear reactor control

    DOE Patents [OSTI]

    Cawley, William E.; Warnick, Robert F.

    1982-01-01

    1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

  10. REACTOR VIEWING APPARATUS

    DOE Patents [OSTI]

    Monk, G.S.

    1959-01-13

    An optical system is presented that is suitable for viewing objects in a region of relatively high radioactivity, or high neutron activity, such as a neutronic reactor. This optical system will absorb neutrons and gamma rays thereby protecting personnel fronm the harmful biological effects of such penetrating radiations. The optical system is comprised of a viewing tube having a lens at one end, a transparent solid member at the other end and a transparent aqueous liquid completely filling the tube between the ends. The lens is made of a polymerized organic material and the transparent solid member is made of a radiation absorbent material. A shield surrounds the tube betwcen the flanges and is made of a gamma ray absorbing material.

  11. Neutrino Experiments at Reactors

    DOE R&D Accomplishments [OSTI]

    Reines, F.; Gurr, H. S.; Jenkins, T. L.; Munsee, J. H.

    1968-09-09

    A description is given of the electron-antineutrino program using a large fission reactor. A search has been made for a neutral weak interaction via the reaction (electron antineutrino + d .> p + n + electron antineutrino), the reaction (electron antineutrino + d .> n + n + e{sup +}) has now been detected, and an effort is underway to observe the elastic scattering reaction (electron antineutrino + e{sup -} .> electron antineutrino + e{sup -}) as well as to measure more precisely the reaction (electron antineutrino + p .> n + e{sup+}). The upper limit on the elastic scattering reaction which we have obtained with our large composite NaI, plastic, liquid scintillation detector is now about 50 times the predicted value.

  12. FLUID MODERATED REACTOR

    DOE Patents [OSTI]

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

    1957-10-22

    A reactor which utilizes fissionable fuel elements in rod form immersed in a moderator or heavy water and a means of circulating the heavy water so that it may also function as a coolant to remove the heat generated by the fission of the fuel are described. In this design, the clad fuel elements are held in vertical tubes immersed in heavy water in a tank. The water is circulated in a closed system by entering near the tops of the tubes, passing downward through the tubes over the fuel elements and out into the tank, where it is drawn off at the bottom, passed through heat exchangers to give up its heat and then returned to the tops of the tubes for recirculation.

  13. Nuclear reactor control apparatus

    DOE Patents [OSTI]

    Sridhar, Bettadapur N. (Cupertino, CA)

    1983-10-25

    Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the safety rod to release the rod from the drive shaft. In preferred embodiments of the invention, an additional latch is provided to release a lower portion of a safety rod under conditions that may interfere with movement of the entire rod.

  14. NUCLEAR REACTOR CONTROL SYSTEM

    DOE Patents [OSTI]

    Howard, D.F.; Motta, E.E.

    1961-06-27

    A method for controlling the excess reactivity in a nuclear reactor throughout the core life while maintaining the neutron flux distribution at the desired level is described. The control unit embodies a container having two electrodes of different surface area immersed in an electrolytic solution of a good neutron sbsorbing metal ion such as boron, gadolinium, or cadmium. Initially, the neutron absorber is plated on the larger electrode to control the greater neutron flux of a freshly refueled core. As the fuel burns up, the excess reactivity decreases and the neutron absorber is then plated onto the smaller electrode so that the number of neutrons absorbed also decreases. The excess reactivity in the core may thus be maintained without the introduction of serious perturbations in the neutron flux distributibn.

  15. REACTOR MODERATOR STRUCTURE

    DOE Patents [OSTI]

    Fraas, A.P.; Tudor, J.J.

    1963-08-01

    An improved moderator structure for nuclear reactors consists of moderator blocks arranged in horizontal layers to form a multiplicity of vertically stacked columns of blocks. The blocks in each vertical column are keyed together, and a ceramic grid is disposed between each horizontal layer of blocks. Pressure plates cover- the lateral surface of the moderator structure in abutting relationship with the peripheral terminal lengths of the ceramic grids. Tubular springs are disposed between the pressure plates and a rigid external support. The tubular springs have their axes vertically disposed to facilitate passage of coolant gas through the springs and are spaced apart a selected distance such that at sonae preselected point of spring deflection, the sides of the springs will contact adjacent springs thereby causing a large increase in resistance to further spring deflection. (AEC)

  16. GAS COOLED NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Rodwell, W.

    1958-06-10

    A gas-cooled nuclear reactor consisting of a graphite reacting core and reflector structure supported in a containing vessel is described. A gas sealing means is included for sealing between the walls of the graphite structure and containing vessel to prevent the gas coolant by-passing the reacting core. The reacting core is a multi-sided right prismatic structure having a pair of parallel slots around its periphery. The containing vessel is cylindrical and has a rib on its internal surface which supports two continuous ring shaped flexible web members with their radially innermost ends in sealing engagement within the radially outermost portion of the slots. The core structure is supported on ball bearings. This design permits thermal expansion of the core stracture and vessel while maintainirg a peripheral seal between the tvo elements.

  17. Thermal analysis of a coaxial helium panel of a cryogenic vacuum pump for advanced divertor of DIII-D tokamak

    SciTech Connect (OSTI)

    Baxi, C.B.; Langhorn, A.; Schaubel, K.; Smith, J.

    1991-08-01

    It is planned to install a 50,000 1/s cryogenic pump for particle removal in the D3-D tokamak. A critical component of this cryogenic pump will be a helium panel which has to be maintained at a liquid helium temperature. The outer surface area of the helium panel has an area of 1 m{sup 2} and consists of a 2.5 cm diameter, 10 m long tube. From design considerations, a coaxial geometry is preferable since it requires a minimum number of welds. However, the coaxial geometry also results in a counter flow heat exchanger arrangement, where the outgoing warm fluid will exchange heat with incoming cold fluid. This is of concern since the helium panel must be cooled from liquid nitrogen temperature to liquid helium temperature in less than 5 minutes for successful operation of the cryogenic pump. In order to analyze the thermal performance of the coaxial helium panel, a finite difference computer model of the geometry was prepared. The governing equations took into account axial as well as radial conduction through the tube walls. The variation of thermal properties was modeled. The results of the analysis showed that although the coaxial geometry behaves like a counter flow heat exchanger, within the operating range of the cryogenic pump a rapid cooldown of the helium panel from liquid nitrogen temperature to the operating temperature is feasible. A prototypical experiment was also performed at General Atomics (GA) which verified the concept and the analysis. 4 refs., 8 figs.

  18. Natural Circulation Patterns in the VHTR Air-Ingress Accident and Related Issues

    SciTech Connect (OSTI)

    Chang H. Oh; Eung S. Kim

    2012-08-01

    Natural circulation patterns in the VHTR during a hypothetical air-ingress accident have been investigated using computational fluid dynamic (CFD) methods in order to compare results from the previous 1-D model which was developed using GAMMA code for the air-ingress analyses. The GT-MHR 600 MWt reactor was selected to be the reference design and modeled by a half symmetric 3-D geometry using FLUENT 6.3, a commercial CFD code. CFD simulations were carried out as the steady-state calculation, and the boundary conditions were either assumed or provided from the 1-D GAMMA code results. Totally, 12 different cases have been reviewed, and many notable results have been obtained through in this work. According to the simulations, natural circulation patterns in the reactor were quite different from the previous 1-D assumptions. A large re-circulation flow with thermal stratification phenomena was clearly observed in the hot-leg and the lower plenum in the 3-D model. This re-circulation flow provided about an order faster air-ingress speed (0.46 m/s in superficial velocity) than previously predicted by 1-D modeling (0.02~0.03 m/s). It indicates that the 1-D air-ingress modeling may significantly distort the air-ingress scenario and consequences. In addition, complicated natural circulation patterns are eventually expected to result in very complex graphite oxidations and corrosion behaviors.

  19. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect (OSTI)

    Not Available

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

  20. University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor

    SciTech Connect (OSTI)

    Eric C. Woolstenhulme; Dana M. Hewit

    2008-09-01

    The Department of Energy’s Idaho National Laboratory, under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at Purdue University Reactor. With this work completed and in anticipation of other impending conversion projects, the INL convened and engaged the project participants in a structured discussion to capture the lessons learned. The lessons learned process has allowed us to capture gaps, opportunities, and good practices, drawing from the project team’s experiences. These lessons will be used to raise the standard of excellence, effectiveness, and efficiency in all future conversion projects.