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1

Preliminary safety evaluation of the Gas Turbine-Modular Helium Reactor (GT-MHR)  

SciTech Connect

A qualitative comparison between the safety characteristics of the Gas Turbine-Modular Helium Reactor (GT-MHR) and those of the steam cycle shows that the two designs achieve equivalent levels of overall safety performance. This comparison is obtained by applying the scaling laws to detailed steam-cycle computations as well as the conclusions obtained from preliminary GT-MHR model simulations. The gas turbine design is predicted to be superior for some event categories, while the steam cycle design is better for others. From a safety perspective, the GT-MHR has a modest advantage for pressurized conduction cooldown events. Recent computational simulations of 102 column, 550 MW(t) GT-MHR during a depressurized conduction cooldown show that peak fuel temperatures are within the limits. The GT-MHR has a significantly lower risk due to water ingress events under operating conditions. Two additional scenarios, namely loss of load event and turbine deblading event that are specific to the GT-MHR design are discussed. Preliminary evaluation of the GT-MHR`s safety characteristics indicate that the GT-MHR can be expected to satisfy or exceed its safety requirements.

Dunn, T.D.; Lommers, L.J.; Tangirala, V.E.

1994-04-01T23:59:59.000Z

2

GT-MHR COMMERCIALIZATION STUDY FINAL CONTRACTUAL REPORT OF WORK PERFORMED FROM CONTRACT BEGINNING, JUNE 18,2001,TO CONTRACT END, JANUARY 31,2004  

Science Conference Proceedings (OSTI)

OAK-B135 This is the final report of work performed by General Atomics on a Gas Turbine Modular Helium Reactor (GT-MHR) commercialization study under contract to the Department of Energy, Oakland Operations Office. The contract work scope covered a series of discrete tasks relating to commercialization of the GT-MHR. During the first year of performance, June 18, 2001--June 30, 2002, the contract covered six tasks, Tasks 1 through 6. Subsequently, four additional tasks were added, Tasks 7,8,10 and 11. With the exception of Task 1, each of the contract Tasks involved the development of one or more discrete deliverable products. Task 1 covered activities performed by General Atomics as part of a several year fuel irradiation testing activity being conducted in cooperation with the European Union. The irradiation testing will not be completed for three or more years into the future. Future work by General Atomics on this irradiation test activity will be covered by a new contract.

SHENOY, AS

2004-02-01T23:59:59.000Z

3

Evaluation of Coatings to Prevent Diffusion of Fission Products into Gas Reactor Turbomachine Blades  

Science Conference Proceedings (OSTI)

The defining attributes of the High Temperature Gas-Cooled Reactor (HTGR) (ceramic-based fuel system, graphite moderator, and helium coolant) provide a high temperature capability that is unique among presently demonstrated nuclear energy concepts. In two current project initiatives -- the Gas Turbine Modular Helium Reactor (GT-MHR) and the Pebble Bed Modular Reactor (PBMR) -- the HTGR nuclear heat source is directly coupled to a closed gas turbine (Brayton) cycle, with net generation efficiencies projec...

2004-01-26T23:59:59.000Z

4

GT-MHR power conversion system: Design status and technical issues  

SciTech Connect

The Modular Helium Reactor (MHR) builds on 30 years of international gas-cooled reactor experience utilizing the unique properties of helium gas coolant, graphite moderator and coated particle fuel. To efficiently utilize the high temperature potential of the MHR, an innovative power conversion system has been developed featuring an intercooled and recuperated gas turbine. The gas turbine replaces a conventional steam turbine and its many auxiliary components. The Power Conversion System converts the thermal energy of the helium directly into electrical energy utilizing a closed Brayton cycle. The Power Conversion System draws on even more years of experience than the MHR: the world`s first closed-cycle plant, fossil fired and utilizing air as working fluid, started operation in Switzerland in 1939. Shortly thereafter, in 1945, the coupling of a closed-cycle plant to a nuclear heat generation system was conceived. Directly coupling the closed-cycle gas turbine concept to a modern, passively safe nuclear reactor opens a new chapter in power generation technology and brings with it various design challenges. Some of these challenges are associated with the direct coupling of the Power Conversion System to a nuclear reactor. Since the primary coolant is also the working fluid, the Power Conversion System has to be designed for reactor radionuclide plateout. As a result, issues like component maintainability and replaceability, and fission product effects on materials must be addressed. Other issues concern the integration of the Power Conversion System components into a single vessel. These issues include the selection of key technologies for the power conversion components such as submerged generator, magnetic bearings, seals, compact heat exchangers, and the overall system layout.

Etzel, K.; Baccaglini, G.; Schwartz, A. [General Atomics, San Diego, CA (United States); Hillman, S.; Mathis, D. [AlliedSignal Aerospace, Tempe, AZ (United States)

1994-12-01T23:59:59.000Z

5

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

Science Conference Proceedings (OSTI)

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.

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

2008-10-01T23:59:59.000Z

6

Plateout Phenomena in Direct-Cycle High Temperature Gas-Cooled Reactors  

Science Conference Proceedings (OSTI)

The plateout of condensable radionuclides in the primary coolant circuits of high-temperature gas-cooled reactors (HTGRs) -- particularly direct-cycle HTGRs -- has significant design, operations and maintenance (O&M), and safety implications. This report reviews and evaluates the available international information on plateout phenomena, specifically as it applies to the gas turbine-modular helium reactor (GT-MHR) and the pebble bed modular reactor (PBMR).

2002-06-26T23:59:59.000Z

7

Rotor Scale Model Tests for Power Conversion Unit of GT-MHR (A26845)  

E-Print Network (OSTI)

Proc. Of 5th International Conf. On High Temperatore Reactor Technology, Prague, Czech Republic, 2010; General Atomics Report GA-A26845 (2010)5th International Conference on High Temperature Reactor Technology Prague, CZ, 2010999619072

Baxi, C.B.

2010-08-18T23:59:59.000Z

8

Evaluation of Materials Issues in the PBMR and GT-MHR  

Science Conference Proceedings (OSTI)

Advanced nuclear energy is experiencing a renewed level of interest, both in the United States and internationally. Based on their potential for higher efficiency via the direct gas-turbine (Brayton) cycle and advanced process energy applications such as hydrogen production, high-temperature gas-cooled reactors (HTGRs) are among several advanced nuclear options being considered. In response to this interest, EPRI is assessing several enabling HTGR technologies to determine the state of their development ...

2002-11-18T23:59:59.000Z

9

Prismatic modular reactor analysis with melcor  

E-Print Network (OSTI)

Hydrogen, a more sustainable source of energy, is a potential substitute for hydrocarbon fuel for power generation. The Very High Temperature gas-cooled Reactor (VHTR) concept can produce hydrogen with high efficiency and in large quantities. The US Department of Energy plans to build a VHTR as a next-generation hydrogen/electricity production plant. This reactor concept is very different from that of commercial reactors in the US. In order to acquire licensing eligibility for VHTRs, analysis tools need to be validated and applied to design and evaluate VHTRs under operation conditions and accident scenarios. In this thesis, MELCOR, a severe accident code, was used to analyze one of the VHTR designs – a prismatic core Next Generation Nuclear Plant (NGNP). The NGNP is based on General Atomics‘ (GA) Gas Turbine – Modular Helium Reactor (GT-MHR) 600 MW design. According to the current literature survey, more data is available for the GT-MHR than for the NGNP. Therefore, for the purposes of extending MELCOR capabilities and code validation, a model of the GT-MHR reactor pressure vessel (RPV) was developed. Based on the currently available data, a model of the NGNP RPV was then developed through modifying the GT-MHR RPV model. For both RPV models, coolant outlet temperature under normal operating conditions corresponds well to the data from literature. The reactor cavity cooling systems (RCCS), which passively removes heat from the RPV wall to the outside atmosphere, was then added to this GT-MHR RPV model. With this model addition, the heat removal rate of the RCCS under normal operating conditions was calculated to correspond well to the data from references. Pressurized conduction cooldown (PCC), one of the important postulated accident scenarios for a prismatic core reactor, was simulated with the complete model. MELCOR has been demonstrated to have the ability of modeling a prismatic core VHTR. The calculated outlet temperature and mass flow rate under normal operation correspond well to references. However, the calculation for the heat distribution in the graphite and fuel is unsatisfactory which requires MELCOR modification for the PCC simulation. For future work, a complete model of the NGNP under normal operation conditions will be developed when additional data becomes available.

Zhen, Ni

2008-12-01T23:59:59.000Z

10

Helium turbines for high-temperature reactors  

SciTech Connect

From joint meeting of the VDE and VDI; Dusseldorf, Ger. (17 Oct 1972). The designs of turbines with air and helium as working media are compared, and volume flow, mass flow, sound velocity, stage number, and other characteristics are individually dealt with. Similar comparisons are made regarding connecting lines and heat exchangers. The combination of the helium turbine with hightemperature reactors is described. Problems of the integrated and non- integrated method of single cycle plants and helium turbines, the use of dry cooling towers and the development of helium turbines are discussed. (GE)

Knuefer, H.

1973-12-01T23:59:59.000Z

11

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

DOE Green Energy (OSTI)

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.

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

2003-01-01T23:59:59.000Z

12

Hydrogen Production Using the Modular Helium Reactor  

DOE Green Energy (OSTI)

The high-temperature characteristics of the Modular Helium Reactor (MHR) make it a strong candidate for the production of hydrogen using either thermochemical or high-temperature electrolysis (HTE) processes. Using heat from the MHR to drive a Sulfur-Iodine (S-I) thermochemical hydrogen process has been the subject of a DOE sponsored Nuclear Engineering Research Initiative (NERI) project lead by General Atomics, with participation from the Idaho National Engineering and Environmental Laboratory (INEEL) and Texas A&M University. While the focus of much of the initial work was on the S-I thermochemical production of hydrogen, recent activities have also included development of a preconceptual design for an integral HTE hydrogen production plant driven by the process heat and electricity produced by a 600 MWt MHR. This paper describes RELAP5-3D analyses performed to evaluate alternative primary system cooling configurations for the MHR to minimize peak reactor vessel and core temperatures while achieving core helium outlet temperatures in the range of 900 oC to 1000 oC, needed for the efficient production of hydrogen using either the S-I thermochemical or HTE process. The cooling schemes investigated are intended to ensure peak fuel temperatures do not exceed specified limits under normal or transient upset conditions, and that reactor vessel temperatures do not exceed ASME code limits for steady-state or transient conditions using standard LWR vessel materials. Preconceptual designs for both an S-I thermochemical and HTE hydrogen production plant driven by a 600 MWt MHR at helium outlet temperatures in the range of 900 oC to 1000 oC are described and compared. An initial SAPHIRE model to evaluate the reliability, maintainablility, and availability of the S-I hydrogen production plant is also discussed, and plans for future assessments of conceptual designs for both a S-I thermochemical and HTE hydrogen production plant coupled to a 600 MWt modular helium reactor are described.

E. A. Harvego; S. M. Reza; M. Richards; A. Shenoy

2005-05-01T23:59:59.000Z

13

Liquid uranium alloy-helium fission reactor  

DOE Patents (OSTI)

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.

Minkov, V.

1984-06-13T23:59:59.000Z

14

Liquid uranium alloy-helium fission reactor  

SciTech Connect

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.

Minkov, Vladimir (Skokie, IL)

1986-01-01T23:59:59.000Z

15

Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel  

DOE Green Energy (OSTI)

The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [{approx}1000 Mw(e)] with passive safety systems to provide the potential for improved economics.

Forsberg, C.W.

2002-02-21T23:59:59.000Z

16

STATEMENT OF CONSIDERATIONS REQUEST BY GENERAL ATOMICS FOR AN...  

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

and Selection Of GA Prior to 1995, DOE funded GA's work to develop technology related to Gas Turbine Modular Helium Reactors (GT-MHR). See Appendix A, Petition Answers 4 and 8....

17

Computational Fluid Dynamics Analyses on Very High Temperature Reactor Air Ingress  

SciTech Connect

A preliminary computational fluid dynamics (CFD) analysis was performed to understand density-gradient-induced stratified flow in a Very High Temperature Reactor (VHTR) air-ingress accident. Various parameters were taken into consideration, including turbulence model, core temperature, initial air mole-fraction, and flow resistance in the core. The gas turbine modular helium reactor (GT-MHR) 600 MWt was selected as the reference reactor and it was simplified to be 2-D geometry in modeling. The core and the lower plenum were assumed to be porous bodies. Following the preliminary CFD results, the analysis of the air-ingress accident has been performed by two different codes: GAMMA code (system analysis code, Oh et al. 2006) and FLUENT CFD code (Fluent 2007). Eventually, the analysis results showed that the actual onset time of natural convection (~160 sec) would be significantly earlier than the previous predictions (~150 hours) calculated based on the molecular diffusion air-ingress mechanism. This leads to the conclusion that the consequences of this accident will be much more serious than previously expected.

Chang H Oh; Eung S. Kim; Richard Schultz; David Petti; Hyung S. Kang

2009-07-01T23:59:59.000Z

18

Fundamental Thermal Fluid Physics of High Temperature Flows in Advanced Reactor Systems - Nuclear Energy Research Initiative Program Interoffice Work Order (IWO) MSF99-0254 Final Report for Period 1 August 1999 to 31 December 2002  

DOE Green Energy (OSTI)

The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of advanced reactors for higher efficiency and enhanced safety and for deployable reactors for electrical power generation, process heat utilization and hydrogen generation. While key applications would be advanced gas-cooled reactors (AGCRs) using the closed Brayton cycle (CBC) for higher efficiency (such as the proposed Gas Turbine - Modular Helium Reactor (GT-MHR) of General Atomics [Neylan and Simon, 1996]), results of the proposed research should also be valuable in reactor systems with supercritical flow or superheated vapors, e.g., steam. Higher efficiency leads to lower cost/kwh and reduces life-cycle impacts of radioactive waste (by reducing waters/kwh). The outcome will also be useful for some space power and propulsion concepts and for some fusion reactor concepts as side benefits, but they are not the thrusts of the investigation. The objective of the project is to provide fundamental thermal fluid physics knowledge and measurements necessary for the development of the improved methods for the applications.

McEligot, D.M.; Condie, K.G.; Foust, T.D.; McCreery, G.E.; Pink, R.J.; Stacey, D.E. (INEEL); Shenoy, A.; Baccaglini, G. (General Atomics); Pletcher, R.H. (Iowa State U.); Wallace, J.M.; Vukoslavcevic, P. (U. Maryland); Jackson, J.D. (U. Manchester, UK); Kunugi, T. (Kyoto U., Japan); Satake, S.-i. (Tokyo U. Science, Japan)

2002-12-31T23:59:59.000Z

19

EPRI NMAC Maintainability Review of the International Gas-Turbine Modular Helium Reactor Power Conversion Unit  

Science Conference Proceedings (OSTI)

This report provides information of interest to the designers of modular helium-reactor-driven gas turbines and persons considering the purchase of this type of plant.

2001-02-01T23:59:59.000Z

20

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

DOE Green Energy (OSTI)

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 both the PBMR and prismatic design. The main focus of this report is the RPV for both design concepts with emphasis on material selection.

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

2007-03-21T23:59:59.000Z

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


21

USE OF THE MODULAR HELIUM REACTOR FOR HYDROGEN PRODUCTION  

DOE Green Energy (OSTI)

OAK-B135 A significant ''Hydrogen Economy'' is predicted that will reduce our dependence on petroleum imports and reduce pollution and greenhouse gas emissions. Hydrogen is an environmentally attractive fuel that has the potential to displace fossil fuels, but contemporary hydrogen production is primarily based on fossil fuels. The author has recently completed a three-year project for the US Department of Energy (DOE) whose objective was to ''define an economically feasible concept for production of hydrogen, using an advanced high-temperature nuclear reactor as the energy source''. Thermochemical water-slitting, a chemical process that accomplishes the decomposition of water into hydrogen and oxygen, met this objective. The goal of the first phase of this study was to evaluate thermochemical processes which offer the potential for efficient, cost-effective, large-scale production of hydrogen, and to select one for further detailed consideration. They selected the Sulfur-Iodine cycle. In the second phase, they reviewed all the basic reactor types for suitability to provide the high temperature heat needed by the selected thermochemical water splitting cycle and chose the helium gas-cooled reactor. In the third phase they designed the chemical flowsheet for the thermochemical process and estimated the efficiency and cost of the process and the projected cost of producing hydrogen. These results are summarized in this report.

SCHULTZ,KR

2003-09-01T23:59:59.000Z

22

Thermal and flow design of helium-cooled reactors  

Science Conference Proceedings (OSTI)

This book continues the American Nuclear Society's series of monographs on nuclear science and technology. Chapters of the book include information on the first-generation gas-cooled reactors; HTGR reactor developments; reactor core heat transfer; mechanical problems related to the primary coolant circuit; HTGR design bases; core thermal design; gas turbines; process heat HTGR reactors; GCFR reactor thermal hydraulics; and gas cooling of fusion reactors.

Melese, G.; Katz, R.

1984-01-01T23:59:59.000Z

23

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

Science Conference Proceedings (OSTI)

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 this report, namely the development of guidance to support the consideration of MC&A in the design of both pebble-bed and prismatic-fueled HTGRs. The objective is to identify and incorporate design features into the facility design that will cost effectively aid in making MC&A more effective and efficient, with minimum impact on operations. The theft of nuclear material is addressed through both MC&A and physical protection, while the threat of sabotage is addressed principally through physical protection.

Trond Bjornard; John Hockert

2011-08-01T23:59:59.000Z

24

Main features of direct cycle helium gas turbines integrated with a high temperature reactor  

SciTech Connect

From international nuclear industries fair; Basel, Switzerland (16 Oct 1972). The main features and advantages of direct cycle helium gas turbines integrated with a high temperature reactor are presented. The proposed design concept is based on a logical extension of existirg knowledge and experience on currently built gas cooled reactors and industrial gas turbines. The direct cycle gas turbine offers many advantages in the form of high reliability, safety and simplicity; it emerges as a potential competitor to the main power generation prime mover, the steam turbine. (auth)

Burylo, P.

1972-01-01T23:59:59.000Z

25

A helium-cooled blanket design of the low aspect ratio reactor  

Science Conference Proceedings (OSTI)

An aggressive low aspect ratio scoping fusion reactor design indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of > 4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 436 C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m(2), respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650 C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh.

Wong, C.P.; Baxi, C.B.; Reis, E.E. [General Atomics, San Diego, CA (United States); Cerbone, R.; Cheng, E.T. [TSI Research, Solana Beach, CA (United States)

1998-03-01T23:59:59.000Z

26

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

SciTech Connect

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.

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

2010-08-01T23:59:59.000Z

27

Plate-Out Phenomena in Direct-Cycle HTGRs  

Science Conference Proceedings (OSTI)

The plateout of condensable radionuclides in the primary coolant circuits of high-temperature gas-cooled reactors (HTGRs) -- especially direct-cycle HTGRs -- has significant design, operations and maintenance (O&M), and safety implications. This report reviews and evaluates the available international information on plateout phenomena, specifically as it applies to the gas turbine-modular helium reactor (GT-MHR) and the pebble bed modular reactor (PBMR).

2001-12-05T23:59:59.000Z

28

STUDY ON AIR INGRESS MITIGATION METHODS IN THE VERY HIGH TEMPERATURE GAS COOLED REACTOR (VHTR)  

SciTech Connect

An air-ingress accident followed by a pipe break is considered as a critical event for a very high temperature gas-cooled reactor (VHTR). Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break leading to oxidation of the in-core graphite structure. Thus, without mitigation features, this accident might lead to severe exothermic chemical reactions of graphite and oxygen. Under extreme circumstances, a loss of core structural integrity may occur along with excessive release of radiological inventory. Idaho National Laboratory under the auspices of the U.S. Department of Energy is performing research and development (R&D) that focuses on key phenomena important during challenging scenarios that may occur in the VHTR. Phenomena Identification and Ranking Table (PIRT) studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Oh et al. 2006, Schultz et al. 2006). Consequently, the development of advanced air ingress-related models and verification and validation (V&V) requirements are part of the experimental validation plan. This paper discusses about various air-ingress mitigation concepts applicable for the VHTRs. The study begins with identifying important factors (or phenomena) associated with the air-ingress accident by using a root-cause analysis. By preventing main causes of the important events identified in the root-cause diagram, the basic air-ingress mitigation ideas can be conceptually derived. The main concepts include (1) preventing structural degradation of graphite supporters; (2) preventing local stress concentration in the supporter; (3) preventing graphite oxidation; (4) preventing air ingress; (5) preventing density gradient driven flow; (4) preventing fluid density gradient; (5) preventing fluid temperature gradient; (6) preventing high temperature. Based on the basic concepts listed above, various air-ingress mitigation methods are proposed in this study. Among them, the following two mitigation ideas are extensively investigated using computational fluid dynamic codes (CFD): (1) helium injection in the lower plenum, and (2) reactor enclosure opened at the bottom. The main idea of the helium injection method is to replace air in the core and the lower plenum upper part by buoyancy force. This method reduces graphite oxidation damage in the severe locations of the reactor inside. To validate this method, CFD simulations are addressed here. A simple 2-D CFD model is developed based on the GT-MHR 600MWt design. The simulation results showed that the helium replace the air flow into the core and significantly reduce the air concentration in the core and bottom reflector potentially protecting oxidation damage. According to the simulation results, even small helium flow was sufficient to remove air in the core, mitigating the air-ingress successfully. The idea of the reactor enclosure with an opening at the bottom changes overall air-ingress mechanism from natural convection to molecular diffusion. This method can be applied to the current system by some design modification of the reactor cavity. To validate this concept, this study also uses CFD simulations based on the simplified 2-D geometry. The simulation results showed that the enclosure open at the bottom can successfully mitigate air-ingress into the reactor even after on-set natural circulation occurs.

Chang H. Oh

2011-03-01T23:59:59.000Z

29

Assessment of passive decay heat removal in the General Atomics Modular Helium Reactor  

E-Print Network (OSTI)

The purpose of this report is to present the results of the study and analysis of loss-of-coolant and loss-of-flow simulations performed on the Modular Helium Reactor developed by General Atomics using the thermal-hydraulics code RELAP5-3D/ATHENA. The MHR is a high temperature gas cooled reactor. It is a prismatic core concept for New Generation Nuclear Plant (NGNP). Very few reactors of that kind have been designed in the past. Furthermore, the MHR is supposed to be a highly passively safe concept. So there are high needs for numerical simulations in order to confirm the design. The project is dedicated to the assessment of the passive decay heat capabilities of the reactor under abnormal transient conditions. To comply with the requirements of the NGNP, fuel and structural temperatures must be kept under design safety limits under any circumstances. During the project, the MHR has been investigated: first under steady-state conditions and then under transient settings. The project confirms that satisfying passive decay heat removal by means of natural heat transfer mechanisms (convection, conduction and radiation) occurs.

Cocheme, Francois Guilhem

2004-12-01T23:59:59.000Z

30

Lifetime Test of a Partial Model of a High-Temperature Gas-Cooled Reactor Helium-Helium Heat Exchanger  

Science Conference Proceedings (OSTI)

H. Design Codes and Life Prediction / Status of Metallic Materials Development for Application in Advanced High-Temperature Gas-Cooled Reactor / Material

Masaki Kitagawa; Hiroshi Hattori; Akira Ohtomo; Tetsuo Teramae; Junichi Hamanaka; Hiroshi Ukikusa

31

Fuel Cycle Optimization of a Helium-Cooled, Sub-Critical, Fast Transmutation of Waste Reactor with a Fusion Neutron Source.  

E-Print Network (OSTI)

??Possible fuel cycle scenarios for a helium-cooled, sub-critical, fast reactor with a fusion neutron source for the transmutation of spent nuclear fuel have been analyzed.… (more)

Maddox, James Warren

2006-01-01T23:59:59.000Z

32

Helium turbine design for a 1000 MWe gas-cooled fast breeder reactor with closed gas turbine cycle  

SciTech Connect

This report deals exclusively with the preliminary design of a double-flooded helium turbine for a 1000 MWe gas-cooled fast breeder reactor. The influence is studied of several parameters, such as hub ratio, exit angle of the turbine wheel and inlet angle of the guide wheel, on the designed size of the turbine and the centrifugal stress of the blading, in order to get a survey which is helpful in the preliminary design.

Savatteri, C.

1973-02-15T23:59:59.000Z

33

High-temperature gas-cooled reactor helium compatibility studies: results of 10,000-hour exposure of selected alloys in simulated reactor helium  

SciTech Connect

Work on the HTGR Helium Compatibility Task accomplished during the period March 31, 1977 through September 30, 1979, is documented in this report. Emphasis is on the results and analyses of creep data to 10,000 h and the detailed metallurgical evaluations performed on candidate alloy specimens tested for up to 10,000 h. Long-term creep and unstressed aging data in controlled-impurity helium and in air at 800, 900, and 1000/sup 0/C are reported for alloys included in the program in FY-76, including the wrought solid-solution-strengthened alloys, Hastelloy X, Hastelloy S, RA 333, and HD 556, and the centrifugally cast austenitic alloys, HK 40, Supertherm, Manaurite 36X, Manaurite 36XS, and Manaurite 900.

Lechtenberg, T.A.; Stevenson, R.D.; Johnson, W.R.

1980-05-01T23:59:59.000Z

34

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

SciTech Connect

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).

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

2010-10-01T23:59:59.000Z

35

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

Science Conference Proceedings (OSTI)

The DB Program monthly highlights report for September 2010, ORNL/TM-2010/252, was distributed to program participants by email on October 26. This report discusses: (1) Core and Fuel Analysis; (2) Spent Fuel Management; (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor); (4) TRU (transuranic elements) HTR Fuel Qualification; (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.

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

2010-11-01T23:59:59.000Z

36

Generation and Retention of Helium and Hydrogen in Austenitic Steels Irradiated in a Variety of LWR and Test Reactor Spectral Environments  

DOE Green Energy (OSTI)

In fission and fusion reactor environments stainless steels generate significant amounts of helium and hydrogen by transmutation. The primary sources of helium are boron and nickel, interacting with both fast and especially thermal neutrons. Hydrogen arises primarily from fast neutron reactions, but is also introduced into steels at often much higher levels by other environmental processes. Although essentially all of the helium is retained in the steel, it is commonly assumed that most of the hydrogen is not retained. It now appears that under some circumstances, significant levels of hydrogen can be retained, especially when helium-nucleated cavities become a significant part of the microstructure. A variety of stainless steel specimens have been examined from various test reactors, PWRs and BWRs. These specimens were exposed to a wide range of neutron spectra with different thermal/fast neutron ratios. Pure nickel and pure iron have also been examined. It is shown that all major features of the retention of helium and hydrogen can be explained in terms of the composition, thermal/fast neutron ratio and the presence or absence of helium-nucleated cavities. In some cases, the hydrogen retention is very large and can exceed that generated by transmutation, with the additional hydrogen arising from either environmental sources and/or previously unidentified radioisotope sources that may come into operation at high neutron exposures.

Garner, Francis A.; Oliver, Brian M.; Greenwood, Lawrence R.; Edwards, Danny J.; Bruemmer, Stephen M.; Grossbeck, Martin L.

2002-03-31T23:59:59.000Z

37

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

Science Conference Proceedings (OSTI)

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.

Crosswait, K.M.

1994-04-01T23:59:59.000Z

38

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

Science Conference Proceedings (OSTI)

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

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

1990-11-01T23:59:59.000Z

39

The Framatome ANP Indirect-Cycle Very High Temperature Reactor  

SciTech Connect

Framatome ANP is developing a Very High Temperature Reactor (VHTR) design, relying on its previous experience with high temperature reactor designs, from its participation in the MODUL and the GT-MHR designs. The Framatome ANP VHTR design is based on an indirect cycle coupled to an 'off-the-shelf' combined cycle gas turbine. Although direct cycle HTR's are being promoted for their high efficiency, preliminary evaluations show that the Framatome ANP design efficiency is on par with a direct cycle while avoiding PGS (Power Generation System) developments and keeping the PGS contamination free. This concept was independently evaluated with sensitivity analysis by EDF. Moreover, the nuclear heat source of the indirect cycle could also be used to qualify the direct cycle components without risk of contamination behind the IHX, thus assisting in the preparation for the later introduction of that technology. Relying to the maximum extent on available technology, the Framatome ANP VHTR plant can demonstrate high-efficiency electricity generation and carbon-free hydrogen production. (authors)

Copsey, Bernie [Framatome ANP, Inc., 3315 Old Forest Road Lynchburg, VA (United States); Lecomte, Michel [Framatome ANP, SAS, Tour AREVA Paris, La Defense (France); Brinkmann, Gerd [Framatome ANP, GmbH, 49 (9131) 18-96630, Erlangen (Germany); Capitaine, Alain; Deberne, Nicolas [EDF/SEPTEN, Villeurbanne (France)

2004-07-01T23:59:59.000Z

40

Potential applications of helium-cooled high-temperature reactors to process heat use  

DOE Green Energy (OSTI)

High-Temperature Gas-Cooled Reactors (HTRs) permit nuclear energy to be applied to a number of processes presently utilizing fossil fuels. Promising applications of HTRs involve cogeneration, thermal energy transport using molten salt systems, steam reforming of methane for production of chemicals, coal and oil shale liquefaction or gasification, and - in the longer term - energy transport using a chemical heat pipe. Further, HTRs might be used in the more distant future as the energy source for thermochemical hydrogen production from water. Preliminary results of ongoing studies indicate that the potential market for Process Heat HTRs by the year 2020 is about 150 to 250 GW(t) for process heat/cogeneration application, plus approximately 150 to 300 GW(t) for application to fossil conversion processes. HTR cogeneration plants appear attractive in the near term for new industrial plants using large amounts of process heat, possibly for present industrial plants in conjunction with molten-salt energy distribution systems, and also for some fossil conversion processes. HTR reformer systems will take longer to develop, but are applicable to chemicals production, a larger number of fossil conversion processes, and to chemical heat pipes.

Gambill, W.R.; Kasten, P.R.

1981-01-01T23:59:59.000Z

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


41

Design modification for the modular helium reactor for higher temperature operation and reliability studies for nuclear hydrogen production processes  

E-Print Network (OSTI)

Design options have been evaluated for the Modular Helium Reactor (MHR) for higher temperature operation. An alternative configuration for the MHR coolant inlet flow path is developed to reduce the peak vessel temperature (PVT). The coolant inlet path is shifted from the annular path between reactor core barrel and vessel wall through the permanent side reflector (PSR). The number and dimensions of coolant holes are varied to optimize the pressure drop, the inlet velocity, and the percentage of graphite removed from the PSR to create this inlet path. With the removal of ~10% of the graphite from PSR the PVT is reduced from 541 0C to 421 0C. A new design for the graphite block core has been evaluated and optimized to reduce the inlet coolant temperature with the aim of further reduction of PVT. The dimensions and number of fuel rods and coolant holes, and the triangular pitch have been changed and optimized. Different packing fractions for the new core design have been used to conserve the number of fuel particles. Thermal properties for the fuel elements are calculated and incorporated into these analyses. The inlet temperature, mass flow and bypass flow are optimized to limit the peak fuel temperature (PFT) within an acceptable range. Using both of these modifications together, the PVT is reduced to ~350 0C while keeping the outlet temperature at 950 0C and maintaining the PFT within acceptable limits. The vessel and fuel temperatures during low pressure conduction cooldown and high pressure conduction cooldown transients are found to be well below the design limits. The reliability and availability studies for coupled nuclear hydrogen production processes based on the sulfur iodine thermochemical process and high temperature electrolysis process have been accomplished. The fault tree models for both these processes are developed. Using information obtained on system configuration, component failure probability, component repair time and system operating modes and conditions, the system reliability and availability are assessed. Required redundancies are made to improve system reliability and to optimize the plant design for economic performance. The failure rates and outage factors of both processes are found to be well below the maximum acceptable range.

Reza, S.M. Mohsin

2007-05-01T23:59:59.000Z

42

Plant Design and Cost Estimation of a Natural Circulation Lead-Bismuth Reactor with Helium Power Conversion Cycle  

E-Print Network (OSTI)

The analysis of an indirect helium power conversion system with lead-bismuth natural circulation primary system has been performed. The work of this report is focused on 1) identifying the allowable design region for the ...

Kim, D.

43

Effects of a range of machined and ground surface finishes on the simulated reactor helium corrosion of several candidate structural materials. [Inconel MA 754  

SciTech Connect

This report discusses the corrosion behavior of several candidate reactor structural alloys in a simulated advanced high-temperature gas-cooled reactor (HTGR) environment over a range of lathe-machined and centerless-ground surface finishes. The helium environment contained 50 Pa H/sub 2//5 Pa CO/5 Pa CH/sub 4//<0.05 Pa H/sub 2/O (500 ..mu..atm H/sub 2//50 ..mu..atm CO/50 ..mu..atm CH/sub 4//<0.5 ..mu..atm H/sub 2/O) at 900/sup 0/C for a total exposure of 3000 h. The test alloys included two vacuum-cast superalloys (IN 100 and IN 713LC); a centrifugally cast austenitic alloy (HK 40); three wrought high-temperature alloys (Alloy 800H, Hastelloy X, and Inconel 617); and a nickel-base oxide-dispersion-strengthened alloy (Inconel MA 754). Surface finish variations did not affect the simulated advanced-HTGR corrosion behavior of these materials. Under these conditions, the availability of reactant gaseous impurities controls the kinetics of the observed gas-metal interactions. Variations in the near-surface activities and mobilities of reactive solute elements, such as chromium, which might be expected to be affected by changes in surface finish, do not seem to greatly influence corrosion in this simulated advanced HTGR environment. 18 figures, 4 tables.

Thompson, L.D.

1981-02-01T23:59:59.000Z

44

Near-field characterization of hydrogen and helium operation on the TFTR (Tokamak Fusion Test Reactor) diagnostic neutral beam  

DOE Green Energy (OSTI)

An Optical Multichannel Analyzer has been used to measure beam divergence and composition. This measurement is usually performed near the center of the neutralizer or beyond the magnet. In the past, these locations suffered difficult beam composition analysis and low light intensity, respectively. It has been determined that the light emission is relatively independent of neutralizer line density in the near field, allowing near-field measurements to overcome both difficulties. At optimum perveance, but under conditions of high gas throughput, the helium 1/e-divergence angle was measured to be 1.5{degree}. Further investigation found that the divergence decreased with gas throughput down to 1.25{degree}. Mimimum divergences for the full-, half-, and third-energy hydrogen components were 1.1{degree}, 1.2{degree}, and 1.4{degree}, respectively. Relative neutral hydrogen particle fluxes available for injection into TFTR are a function of perveance. At maximum perveance, the full-, half-, and third-energy atom fractions were 0.25 {plus minus} 0.04, 0.5 {plus minus} 0.04, and 0.25 {plus minus} 0.05, respectively. 10 refs., 5 figs.

Kamperschroer, J.H.; Schilling, G.; Roquemore, A.L.

1990-07-01T23:59:59.000Z

45

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

SciTech Connect

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)

Jiang, J.; Yuan, B.; Jin, M.; Wang, M.; Long, P.; Hu, L. [Inst. of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Univ. of Science and Technology of China, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, Univ. of Science and Technology of China, No.350 Shushanhu Road, Shushan District, Hefei, Anhui, 230031 (China)

2012-07-01T23:59:59.000Z

46

Assessment of General Atomics accelerator transmutation of waste concept based on gas-turbine-modular helium cooled reactor technology.  

Science Conference Proceedings (OSTI)

An assessment has been performed for an Accelerator Transmutation of Waste (ATW) concept based on the use of the high temperature gas reactor technology. The concept has been proposed by General Atomics for the ATW system. The assessment was jointly conducted at Argonne National Laboratory (ANL) and Los Alamos national laboratory to assess and to define the potential candidates for the ATW system. This report represents the assessment work performed at ANL. The concept uses recycled light water reactor (LWR)-discharge-transuranic extracted from irradiated oxide fuel in a critical and sub-critical accelerator driven gas-cooled transmuter. In this concept, the transmuter operates at 600 MWt first in the critical mode for three cycles and then operates in a subcritical accelerator-driven mode for a single cycle. The transmuter contains both thermal and fast spectrum transmutation zones. The thermal zone is fueled with the TRU oxide material in the form of coated particles, which are mixed with graphite powder, packed into cylindrical compacts, and loaded in hexagonal graphite blocks with cylindrical channels; the fast zone is fueled with TRU-oxide material in the form of coated particles without the graphite powder and the graphite blocks that has been burned in the thermal region for three critical cycles and one additional accelerator-driven cycle. The fuel loaded into the fast zone is irradiated for four additional cycles. This fuel management scheme is intended to achieve a high Pu isotopes consumption in the thermal spectrum zone, and to consume the minor actinides in the fast-spectrum zone. Monte Carlo and deterministic codes have been used to assess the system performance and to determine the feasibility of achieving high TRU consumption levels. The studies revealed the potential for high consumption of Pu-239 (97%), total Pu (71%) and total TRU (64%) in the system. The analyses confirmed the need for burnable absorber for both suppressing the initial excess reactivity and ensuring a negative temperature coefficient under all operating conditions. Additionally, current results suggest that it may be preferable to use a double strata thermal critical system and fast subcritical system to achieve nearly complete destruction of the TRU oxide fuel. The report gives a general description of the system proposed by General Atomics. The major design parameters (degrees of freedom), which can be altered to optimize the system design, and the constraints, which guide the design and the optimization studies are described. The deterministic and the Monte Carlo neutronics codes and models used for the neutronics analysis and assessment are presented. The results of fuel block and whole-core parametric studies performed to understand the physics are given including the effect of various fuel management schemes on the system performance. A point design is described including the system performance results for a single-batch and three-batch loading schemes. The major design issues, which need to be addressed during further studies, are discussed.

Gohar, Y.; Taiwo, T. A.; Cahalan, J. E.; Finck, P. J.

2001-05-08T23:59:59.000Z

47

STATEMENT OF CONSIDERATIONS REQUEST BY GENERAL ATOMICS FOR AN ADVANCE WAIVER OF THE  

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

GENERAL ATOMICS FOR AN ADVANCE WAIVER OF THE GENERAL ATOMICS FOR AN ADVANCE WAIVER OF THE GOVERNMENT'S DOMESTIC AND FOREIGN PATENT AND COPYRIGHT RIGHTS UNDER DOE CONTRACT DE-AC03-00SF21868; DOE WAIVER NO W(A)-00-017; SAN 685 The Petitioner, General Atomics (GA), has requested an Advance Waiver of the Government's domestic and foreign rights to inventions in the above cited research and development contract (GA Contract). See Appendix A- General Atomics' Petition, Petition Answer I. In addition, GA would like to assert copyright in computer software and delay the immediate release of unpublished technical data. The Development of the Technology and Selection Of GA Prior to 1995, DOE funded GA's work to develop technology related to Gas Turbine Modular Helium Reactors (GT-MHR). See Appendix A, Petition Answers 4 and 8. After the Government

48

Topical report : CFD analysis for the applicability of the natural convection shutdown heat removal test facility (NSTF) for the simulation of the VHTR RCCS.  

DOE Green Energy (OSTI)

The Very High Temperature gas cooled reactor (VHTR) is one of the GEN IV reactor concepts that have been proposed for thermochemical hydrogen production and other process-heat applications like coal gasification. The United States Department of Energy has selected the VHTR for further research and development, aiming to demonstrate emissions-free electricity and hydrogen production at a future time. One of the major safety advantages of the VHTR is the potential for passive decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-side of the RCCS is very similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that has been proposed for the PRISM reactor design. The design and safety analysis of the RVACS have been based on extensive analytical and experimental work performed at ANL. The Natural Convection Shutdown Heat Removal Test Facility (NSTF) at ANL that simulates at full scale the air-side of the RVACS was built to provide experimental support for the design and analysis of the PRISM RVACS system. The objective of this work is to demonstrate that the NSTF facility can be used to generate RCCS experimental data: to validate CFD and systems codes for the analysis of the RCCS; and to support the design and safety analysis of the RCCS. At this time no reference design is available for the NGNP. The General Atomics (GA) gas turbine - modular helium reactor (GT-MHR) has been used in many analyses as a starting reference design. In the GT-MHR the reactor outlet temperature is 850 C, while the target outlet reactor temperature in VHTR is 1000 C. VHTR scoping studies with a reactor outlet temperature of 1000 C have been performed at GA and INEL. Although the reactor outlet temperature in the VHTR is significantly higher than in the GT-MHR, the peak temperature in the reactor vessel (which is the heat source for the RCCS) is not drastically different. In this work, analyses have been performed using reactor vessel temperatures from the GT-MHR design, and the VHTR scoping studies. To demonstrate the applicability of the NSTF facility for full scale simulation of the RCCS the following approach was used. CFD analyses were performed of the RCCS and of its simulation at NSTF to demonstrate that: all significant fluid flow and heat transfer phenomena in the RCCS can be simulated at NSTF; and RCCS simulations at NSTF can cover the whole range of variation of the parameters describing these important phenomena in the RCCS. In CFD analyses, the simulation of turbulence is one of the most significant challenges. Direct Numerical Simulation (DNS) and Large Eddy Simulation (LES) of turbulence in large scale systems require excessive computational resources. The use of the Low-Re number k-{var_epsilon} model, which resolves the boundary layer, is computationally expensive in studies where many simulations have to be performed. In Ref. 2 it was shown that in the RCCS, heat transfer coefficient predictions of the high-Re number k-{var_epsilon} model are closer to those of the low-Re number model than those of heat transfer correlations. In this work, the standard high-Re number k-{var_epsilon} was used to simulate turbulence, and all analyses were performed with the CFD code STARCD.

Tzanos, C. P. (Nuclear Engineering Division)

2007-05-16T23:59:59.000Z

49

Report of the APS Neutrino Study Reactor Working Group  

E-Print Network (OSTI)

Planetary and protostellar nuclear fission: implications forDeep-earth reactor: nuclear fission, helium, and theRaghavan, "Detecting a Nuclear Fission Reactor at the Center

2004-01-01T23:59:59.000Z

50

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

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

is that a permeable wall is used to separate out helium produced as ash by a burning fusion reactor. This would replace the divertor structure and associated pumps in a...

51

REACTOR  

DOE Patents (OSTI)

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

Roman, W.G.

1961-06-27T23:59:59.000Z

52

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

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

1959-10-27T23:59:59.000Z

53

Neutronic reactor thermal shield  

DOE Patents (OSTI)

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.

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

1976-06-15T23:59:59.000Z

54

ISOTHERMAL AIR-INGRESS VALIDATION EXPERIMENTS  

Science Conference Proceedings (OSTI)

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.

Chang H. Oh; Eung S. Kim

2013-01-01T23:59:59.000Z

55

It's Elemental - The Element Helium  

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

to the atmosphere through cracks in the crust. Helium is commercially recovered from natural gas deposits, mostly from Texas, Oklahoma and Kansas. Helium gas is used to inflate...

56

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

DOE Green Energy (OSTI)

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.

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

2006-04-19T23:59:59.000Z

57

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  

DOE Green Energy (OSTI)

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.

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

2006-04-19T23:59:59.000Z

58

Modularity of the MIT Pebble Bed Reactor for use by the commercial power industry  

E-Print Network (OSTI)

The Modular Pebble Bed Reactor is a small high temperature helium cooled reactor that is being considered for both electric power and hydrogen production. Pebble bed reactors are being developed in South Africa, China and ...

Hanlon-Hyssong, Jaime E

2008-01-01T23:59:59.000Z

59

Program for alloy development for irradiation performance in fusion reactors  

SciTech Connect

The use of fission reactors as irradiation test facilities for structural materials for a fusion environment is discussed. A comparison is made of displacement damage and helium production in fast fission and fusion reactors for stainless steel. (MOW)

Stiegler, J.O.; Reuther, T.C.

1977-01-01T23:59:59.000Z

60

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Grebe, J.J.

1959-07-14T23:59:59.000Z

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


61

Helium dilution refrigeration system  

DOE Patents (OSTI)

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.

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

1988-09-13T23:59:59.000Z

62

ISOTHERMAL AIR INGRESS VALIDATION EXPERIMENTS  

SciTech Connect

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.

Chang H Oh; Eung S Kim

2011-09-01T23:59:59.000Z

63

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

Daniels, F.

1962-12-18T23:59:59.000Z

64

Thermal integration of processes to recover helium  

SciTech Connect

New applications for helium have resulted in significant growth in helium demand in recent years. The primary source of helium in the USA is natural gas reservoirs. Typically natural gas reservoirs that contain helium also contain nitrogen. The presence of nitrogen requires that the heating value of the gas be upgraded using nitrogen rejection units (NRU). Thermal integration of the NRU process with the helium recovery process is described. Further integration of NRU and helium recovery process to produce cold helium gas suitable for use as a feed to a helium liquefier is also described. A key aspect of the integrated process is the use of refrigeration from the NRU to reduce power requirements, and capital cost for the helium recovery process.

Pahade, R.F.; Maloney, J.J.; Fisher, T.F. (Union Carbide Corp., Linde Div., Tonawanda, NY (US))

1988-01-01T23:59:59.000Z

65

Reactor technology: power conversion systems and reactor operation and maintenance  

SciTech Connect

The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He/sup 3/ reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored.

Powell, J.R.

1977-01-01T23:59:59.000Z

66

Pulsed helium ionization detection system  

DOE Patents (OSTI)

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.

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

1985-04-09T23:59:59.000Z

67

Satellite spectra for helium-like titanium. Part II  

Science Conference Proceedings (OSTI)

K/sup ..cap alpha../ x-ray spectra of helium-like titanium, Ti XXI, from Tokamak Fusion Test Reactor (TFTR) plasmas have been observed with a high resolution crystal spectrometer and have been used as a diagnostic of central plasma parameters. The data allow detailed comparison with recent theoretical predictions for the Ti XXI helium-like lines and the associated satellite spectrum in the wavelength range from 2.6000 to 2.6400 A. Improved values for the excitation rate coefficients of the Ti XXI resonance line, the intercombination lines and the forbidden line, and new theoretical results on the wavelengths and transition probabilities for beryllium-like satellites due to transitions of the type 1s/sup 2/ 2lnl' - 1s2p2l'' nl'' with n = 2-4 have been calculated.

Bitter, M.; Hill, K.W.; Zarnstorff, M.; von Goeler, S.; Hulse, R.; Johnson, L.C.; Sauthoff, N.R.; Sesnic, S.; Young, K.M.; Tavernier, M.

1985-08-01T23:59:59.000Z

68

Research on Very High Temperature Gas Reactors  

Science Conference Proceedings (OSTI)

Very high temperature gas reactors are helium-cooled, graphite-moderated advanced reactors that show potential for generating low-cost electricity via gas turbines or cogeneration with process-heat applications. This investigation addresses the development status of advanced coatings for nuclear-fuel particles and high-temperature structural materials and evaluates whether these developments are likely to lead to economically competitive applications of the very high temperature gas reactor concept.

1991-08-08T23:59:59.000Z

69

Precision spectroscopy of the helium atom.  

SciTech Connect

Persistent efforts in both theory and experiment have yielded increasingly precise understanding of the helium atom. Because of its simplicity, the helium atom has long been a testing ground for relativistic and quantum electrodynamic effects in few-body atomic systems theoretically and experimentally. Comparison between theory and experiment of the helium spectroscopy in 1s2p{sup 3}P{sub J} can potentially extract a very precise value of the fine structure constant a. The helium atom can also be used to explore exotic nuclear structures. In this paper, we provide a brief review of the recent advances in precision calculations and measurements of the helium atom.

Hu, S.-M.; Lu, Z.-T.; Yan, Z.-C.; Physics; Univ. of Science and Technology of China; Univ. of Chicago; Univ. of New Brunswick

2009-06-01T23:59:59.000Z

70

GAS COOLED NUCLEAR REACTOR STUDY. Final Report  

SciTech Connect

An investigntion was made of the performance of a gas-cooled reactor, designed to provide a source of high temperature heat to a stream of helium. This reactor, in turn, is used as a source of heat for the air stream in a gas- turbine power plant. The reactor design was predicted primarily on the requirement for transferring a large amount of heat to the helium stream with a pressure drop low enough that it will not represent a major loss of power in the power plant. The mass of uranium e uired far criticality under various circumstances was investigated by multigroup calculations, both on desk calculators and on an IBM-704 machine. The gasturbine power plant perfarmance was studied based on a Studebaker-Packard-designed gas-turbine power plant for the propulsion of destroyer-escort vessels. A small experimental program was carried out to study some effects of helium on graphite and on structural steels. (auth)

Thompson, A.S.

1956-07-31T23:59:59.000Z

71

Radiation effects in materials for fusion reactors  

DOE Green Energy (OSTI)

The 14-MeV neutrons produced in a fusion reactor result in different irradiation damage than the equivalent fluence in a fast breeded reactor, not only because of the higher defect generation rate, but because of the production of significant concentrations of helium and hydrogen. Although no fusion test reactor exists, the effects of combined displacement damage plus helium can be studied in mixed-spectrum fission reactors for alloys containing nickel (e.g., austenitic stainless steels). The presence of helium appears to modify vacancy and interstitial recombination such that microstructural development in alloys differs between the fusion and fission reactor environments. Since mechanical properties of alloys are related to the microstructure, the simultaneous production of helium and displacement damage impacts upon key design properties such as tensile, fatigue, creep, an crack growth. Through an understanding of the basic phenomena occurring during irradiation and the relationships between microstructure and properties, alloys can be tailored to minimize radiation-induced swelling and improve mechanical properties in fusion reactor service.

Scott, J.L.; Grossbeck, M.L.; Maziasz, P.J.

1981-01-01T23:59:59.000Z

72

High-temperature helium-loop facility  

Science Conference Proceedings (OSTI)

The high-temperature helium loop is a facility for materials testing in ultrapure helium gas at high temperatures. The closed loop system is capable of recirculating high-purity helium or helium with controlled impurities. The gas loop maximum operating conditions are as follows: 300 psi pressure, 500 lb/h flow rate, and 2100/sup 0/F temperature. The two test sections can accept samples up to 3.5 in. diameter and 5 ft long. The gas loop is fully instrumented to continuously monitor all parameters of loop operation as well as helium impurities. The loop is fully automated to operate continuously and requires only a daily servicing by a qualified operator to replenish recorder charts and helium makeup gas. Because of its versatility and high degree of parameter control, the helium loop is applicable to many types of materials research. This report describes the test apparatus, operating parameters, peripheral systems, and instrumentation system.

Tokarz, R.D.

1981-09-01T23:59:59.000Z

73

GAS COOLED PEBBLE BED REACTOR FOR A LARGE CENTRAL STATION. Reactor Design and Feasibility Study  

SciTech Connect

An optimum econonic design for a high temperature, helium cooled, central station reactor power plant of about 400 Mw of electric power was determined. The core consists of a randomly packed bed of unclad graphite spheres, approximately one in. in diameter, impregnated with U/sup 233/ and thorium such that a conversion ratio of near unity is achieved. The high temperature helium permits steam conditions, at the turbine throttle, of 1000 deg F and 1450 psia. (auth)

Schock, A.; Bruley, D.F.; Culver, H.N.; Ianni, P.W.; Kaufman, W.F.; Schmidt, R.A.; Supp, R.E.

1957-08-01T23:59:59.000Z

74

In Beam Tests of Implanted Helium Targets  

E-Print Network (OSTI)

Targets consisting of 3,4He implanted into thin aluminum foils (approximately 100, 200 or 600 ug/cm^2) were prepared using intense (a few uA) helium beams at low energy (approximately 20, 40 or 100 keV). Uniformity of the implantation was achieved by a beam raster across a 12 mm diameter tantalum collimator at the rates of 0.1 Hz in the vertical direction and 1 Hz in the horizontal direction. Helium implantation into the very thin (approximately 80-100 ug/cm^2) aluminum foils failed to produce useful targets (with only approximately 10% of the helium retained) due to an under estimation of the range by the code SRIM. The range of low energy helium in aluminum predicted by Northcliffe and Shilling and the NIST online tabulation are observed on the other hand to over estimate the range of low energy helium ions in aluminum. An attempt to increase the amount of helium by implanting a second deeper layer was also carried out, but it did not significantly increase the helium content beyond the blistering limit (approximately 6 x 10^17 helium/cm^2). The implanted targets were bombarded with moderately intense 4He and 16O beams of 50-100 particle nA . Rutherford Back Scattering of 1.0 and 2.5 MeV proton beams and recoil helium from 15.0 MeV oxygen beams were used to study the helium content and profile before, during and after bombardments. We observed the helium content and profile to be very stable even after a prolonged bombardment (up to two days) with moderately intense beams of 16O or 4He. Helium implanted into thin (aluminum) foils is a good choice for thin helium targets needed, for example, for a measurement of the 3he(a,g)7Be reaction and the associated S34 astrophysical cross section factor (S-factor).

J. E. McDonald; R. H. France III; R. A. Jarvis; M. W. Ahmed; M. A. Blackston; Th. Delbar; M. Gai; T. J. Kading; Y. Parpottas; B. A. Perdue; R. M. Prior; D. A. Rubin; M. C. Spraker; J. D. Yeomans; L. Weissman; H. R. Weller; E. L. Wilds Jr; ;; UHartford; GCSU; LNS/UConn; TUNL/Duke; UCL/LLN; Yale; NGCSU

2006-08-16T23:59:59.000Z

75

Assessment of Embrittlement of VHTR Structural Alloys in Impure Helium Environments  

SciTech Connect

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 toughness, creep rupture, and fatigue. Although there are some studies on the effects of impurities in helium coolant on creep rupture and fatigue strength, very little is known about their effects on creep crack initiation and crack growth rate at elevated temperatures.

Crone, Wendy; Cao, Guoping; Sridhara, Kumar

2013-05-31T23:59:59.000Z

76

Iodine Stabilized Helium-Neon Laser  

Science Conference Proceedings (OSTI)

... stabilized lasers or spectral lamps that have been recommended for use by international consensus. The iodine stabilized helium-neon laser is just ...

77

Fermilab Central Helium Liquefier Operations  

SciTech Connect

The Central Helium Liquefier (CHL), in conjunction with 24 satellite refrigerators, supplies refrigeration for the Fermilab superconducting accelerator. Liquid from the CHL is transported in a six kilometer circular transfer line and each satellite withdraws the amount required to boost its refrigeration capacity to the necessary level. Unused liquid is presently returned to compressor suction through a 250 kW calorimeter-heater. A helium gas flow of 1.1 kg/s is supplied to the cold box at 15 bars pressure. The gas flows through a demister and an oil adsorgber before it enters the cold box. In the cold box, manufactured by Koch Process Systems and Sulzer Brothers, Ltd., the gas is first cooled by liquid nitrogen and then the flow is split. Three quarters of the flow is further cooled by a series/parallel combination of three oil bearing turbines and then returned to the low pressure side of the heat exchangers. The return gas is used to cool the remaining high pressure gas which is then expanded to 3.5 bars in a 900 L receiver. From there the fluid is transferred into a distribution box, and then routed to either a collection dewar, directly to the ring, or to a heater. The liquid helium, which is utilized by the satellites to increase their cooling capacity, is warmed to near ambient temperature in the satellite heat exchangers. The satellite compressors return the excess inventory to the CHL via a 20 bar gas header. This gas is injected into the high pressure supply to the cold box. The system is shown.

Hodge, G.A.; Rihel, R.K.; Stone, M.E.; Walker, R.J.; /Fermilab

1983-01-01T23:59:59.000Z

78

Neutron-induced dpa, transmutations, gas production, and helium embrittlement of fusion materials  

E-Print Network (OSTI)

In a fusion reactor materials will be subjected to significant fluxes of high-energy neutrons. As well as causing radiation damage, the neutrons also initiate nuclear reactions leading to changes in the chemical composition of materials (transmutation). Many of these reactions produce gases, particularly helium, which cause additional swelling and embrittlement of materials. This paper investigates, using a combination of neutron-transport and inventory calculations, the variation in displacements per atom (dpa) and helium production levels as a function of position within the high flux regions of a recent conceptual model for the "next-step" fusion device DEMO. Subsequently, the gas production rates are used to provide revised estimates, based on new density-functional-theory results, for the critical component lifetimes associated with the helium-induced grain-boundary embrittlement of materials. The revised estimates give more optimistic projections for the lifetimes of materials in a fusion power plant co...

Gilbert, M R; Nguyen-Manh, D; Zheng, S; Packer, L W; Sublet, J -Ch

2013-01-01T23:59:59.000Z

79

Modernizing helium liquefier G-3  

SciTech Connect

The authors describe the process for modernizing the existing liquefier with minimum alteration of its cooling block, that liquefier being helium expansion-type liquefier G-3, made at the Institute of Physical Problems, Academy of Sciences of the USSR, which operates at a stable output of 40 liters/h. A nitrogen tank and a pistontype expander in the preliminary cooling stages and a throttle in the liquefaction stages are used in G-3. Improving the efficiency of such a cooling cycle is limited by the fact that the optimal parameters of throttle stage of liquefaction do not match with the optimal parameters of the expander in the preliminary cooling stages. Thus, for improving cycle efficiency the pressure in the preliminary stages must be increased but reduced in the liquefaction stage. This paper also presents the solution to this problem. It is further demonstrated that the use of vapor-liquid expander in the liquefaction stage of helium cooling cycle helps increase the output of the unit by 40% in a relatively simple way.

Golikov, G.E.; Danilov, I.B.

1985-02-01T23:59:59.000Z

80

Focused helium ion beam milling and deposition  

Science Conference Proceedings (OSTI)

The use of a helium ion microscope with an integrated gas injection system for nanofabrication is explored by demonstrating the milling of fine features into single layered graphene and the controlled deposition of tungsten and platinum wires from gaseous ... Keywords: Beam-induced deposition, Focused ion beam, Gas injection system, Graphene, Helium ion microscope

S. A. Boden; Z. Moktadir; D. M. Bagnall; H. Mizuta; H. N. Rutt

2011-08-01T23:59:59.000Z

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


81

A Helium Isotope Perspective On The Dixie Valley, Nevada, Hydrothermal...  

Open Energy Info (EERE)

Helium Isotope Perspective On The Dixie Valley, Nevada, Hydrothermal System Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: A Helium Isotope...

82

Review of Helium Cooling for Fusion Reactor Applications (A23181)  

E-Print Network (OSTI)

To Be Presented At The 5th Int. Symp. On Fusion Nucl. Technol., Rome, Italy, To Be Published In Fusion Eng. And Design5th International Symposium on Fusion Nuclear Technology Rome, IT, 1999932161691

Baxi, C.B.

1999-09-01T23:59:59.000Z

83

Helium-cooled molten-salt fusion breeder  

Science Conference Proceedings (OSTI)

We present a new conceptual design for a fusion reactor blanket that is intended to produce fissile material for fission power plants. Fast fission is suppressed by using beryllium instead of uranium to multiply neutrons. Thermal fission is suppressed by minimizing the fissile inventory. The molten-salt breeding medium (LiF + BeF/sub 2/ + ThF/sub 4/) is circulated through the blanket and to the on-line processing system where /sup 233/U and tritium are continuously removed. Helium cools the blanket and the austenitic steel tubes that contain the molten salt. Austenitic steel was chosen because of its ease of fabrication, adequate radiation-damage lifetime, and low corrosion by molten salt. We estimate that a breeder having 3000 MW of fusion power will produce 6500 kg of /sup 233/U per year. This amount is enough to provide makeup for 20 GWe of light-water reactors per year or twice that many high-temperature gas-cooled reactors or Canadian heavy-water reactors. Safety is enhanced because the afterheat is low and blanket materials do not react with air or water. The fusion breeder based on a pre-MARS tandem mirror is estimated to cost $4.9B or 2.35 times a light-water reactor of the same power. The estimated cost of the /sup 233/U produced is $40/g for fusion plants costing 2.35 times that of a light-water reactor if utility owned or $16/g if government owned.

Moir, R.W.; Lee, J.D.; Fulton, F.J.; Huegel, F.; Neef, W.S. Jr.; Sherwood, A.E.; Berwald, D.H.; Whitley, R.H.; Wong, C.P.C.; Devan, J.H.

1984-12-01T23:59:59.000Z

84

Gas-cooled reactors  

SciTech Connect

Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing.

Schulten, R.; Trauger, D.B.

1976-01-01T23:59:59.000Z

85

EPRI NMAC Maintainability Review of the Pebble Bed Modular Reactor Demonstration Plant  

Science Conference Proceedings (OSTI)

This report provides information to the designers of pebble bed reactor helium-driven gas turbine plants and to others who are considering the purchase of this type of plant.

2002-05-13T23:59:59.000Z

86

Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation  

Science Conference Proceedings (OSTI)

Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W-5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. System study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kWh. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study.

Wong, C.P.C.; Nygren, R.E.; Baxi, C.B.; Fogarty, P.; Ghoniem, N.; Khater, H.; McCarthy, K.; Merrill, B.; Nelson, B.; Reis, E.E.; Sharafat, S.; Schleicher, R.; Sze, D.K.; Ulrickson, M.; Willms, S.; Youssef, M.; Zinkel, S.

1999-08-01T23:59:59.000Z

87

Compound cryopump for fusion reactors  

E-Print Network (OSTI)

We reconsider an old idea: a three-stage compound cryopump for use in fusion reactors such as DEMO. The helium "ash" is adsorbed on a 4.5 K charcoal-coated surface, while deuterium and tritium are adsorbed at 15-22 K on a second charcoal-coated surface. The helium is released by raising the first surface to ~30 K. In a separate regeneration step, deuterium and tritium are released at ~110 K. In this way, the helium can be pre-separated from other species. In the simplest design, all three stages are in the same vessel, with a single valve to close the pump off from the tokamak during regeneration. In an alternative design, the three stages are in separate vessels, connected by valves, allowing the stages to regenerate without interfering with each other. The inclusion of the intermediate stage would not affect the overall pumping speed significantly. The downstream exhaust processing system could be scaled down, as much of the deuterium and tritium could be returned directly to the reactor. This could reduce ...

Kovari, M; Shephard, T

2013-01-01T23:59:59.000Z

88

Assessment of next generation nuclear plant intermediate heat exchanger design.  

DOE Green Energy (OSTI)

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 an assessment on the issues pertaining to the intermediate heat exchanger (IHX) for the NGNP. A detailed thermal hydraulic analysis, using models developed at ANL, was performed to calculate heat transfer, temperature distribution, and pressure drop. Two IHX designs namely, shell and straight tube and compact heat exchangers were considered in an earlier assessment. Helical coil heat exchangers were analyzed in the current report and the results were compared with the performance features of designs from industry. In addition, a comparative analysis is presented between the shell and straight tube, helical, and printed circuit heat exchangers from the standpoint of heat exchanger volume, primary and secondary sides pressure drop, and number of tubes. The IHX being a high temperature component, probably needs to be designed using ASME Code Section III, Subsection NH, assuming that the IHX will be classified as a class 1 component. With input from thermal hydraulic calculations performed at ANL, thermal conduction and stress analyses were performed for the helical heat exchanger design and the results were compared with earlier-developed results on shell and straight tube and printed circuit heat exchangers.

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

2008-10-17T23:59:59.000Z

89

Nuclear Reactors  

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

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

90

First-principles calculations of the stability and incorporation of helium, xenon and krypton in uranium  

SciTech Connect

While metallic fuels have a long history of reactor use, their fundamental physical and thermodynamic properties are not well understood. Many metallic nuclear fuels are body-centered cubic alloys of uranium that swell under fission conditions, creating fission product gases such as helium, xenon and krypton. In this paper, helium, xenon, and krypton point defects are investigated in the a and ? phases of metallic uranium using first principles calculations. A density functional theory (DFT) framework is utilized with projector augmented-wave (PAW) pseudopotentials. Formation and incorporation energies of He, Xe, and Kr are calculated at various defect positions for the prediction of fission gas behavior in uranium. In most cases, defect energies follow a size effect, with helium incorporation and formation energies being the smallest. The most likely position for the larger Xe and Kr atoms in uranium is the substitutional site. Helium atoms are likely to be found in a wide variety of defect positions due to the comparable formation energies of all defect configurations analyzed. This is the first detailed study of the stability and incorporation of fission gases in uranium.

B. Beeler; B. Good; S. Rashkeev; M. Baskes; M. Okuniewski

2012-06-01T23:59:59.000Z

91

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

DOE R&D Accomplishments (OSTI)

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.

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))

1993-03-00T23:59:59.000Z

92

Neutron-induced dpa, transmutations, gas production, and helium embrittlement of fusion materials  

E-Print Network (OSTI)

In a fusion reactor materials will be subjected to significant fluxes of high-energy neutrons. As well as causing radiation damage, the neutrons also initiate nuclear reactions leading to changes in the chemical composition of materials (transmutation). Many of these reactions produce gases, particularly helium, which cause additional swelling and embrittlement of materials. This paper investigates, using a combination of neutron-transport and inventory calculations, the variation in displacements per atom (dpa) and helium production levels as a function of position within the high flux regions of a recent conceptual model for the "next-step" fusion device DEMO. Subsequently, the gas production rates are used to provide revised estimates, based on new density-functional-theory results, for the critical component lifetimes associated with the helium-induced grain-boundary embrittlement of materials. The revised estimates give more optimistic projections for the lifetimes of materials in a fusion power plant compared to a previous study, while at the same time indicating that helium embrittlement remains one of the most significant factors controlling the structural integrity of fusion power plant components.

M. R. Gilbert; S. L. Dudarev; D. Nguyen-Manh; S. Zheng; L. W. Packer; J. -Ch. Sublet

2013-11-20T23:59:59.000Z

93

Thermal hydraulic design of a 2400 MW t?h? direct supercritical CO?-cooled fast reactor  

E-Print Network (OSTI)

The gas cooled fast reactor (GFR) has received new attention as one of the basic concepts selected by the Generation-IV International Forum (GIF) for further investigation. Currently, the reference GFR is a helium-cooled ...

Pope, Michael A. (Michael Alexander)

2006-01-01T23:59:59.000Z

94

Dynamic Simulation of a Helium Liquefier  

SciTech Connect

Dynamic behavior of a helium liquefier has been studied in detail with a Cryogenic Process REal-time SimulaTor (C-PREST) at the National Institute for Fusion Science (NIFS). The C-PREST is being developed to integrate large-scale helium cryogenic plant design, operation and maintenance for optimum process establishment. As a first step of simulations of cooldown to 4.5 K with the helium liquefier model is conducted, which provides a plant-process validation platform. The helium liquefier consists of seven heat exchangers, a liquid-nitrogen (LN2) precooler, two expansion turbines and a liquid-helium (LHe) reservoir. Process simulations are fulfilled with sequence programs, which were implemented with C-PREST based on an existing liquefier operation. The interactions of a JT valve, a JT-bypass valve and a reservoir-return valve have been dynamically simulated. The paper discusses various aspects of refrigeration process simulation, including its difficulties such as a balance between complexity of the adopted models and CPU time.

Maekawa, R.; Ooba, K.; Mito, T. [National Institute for Fusion Science, Toki, Gifu, 509-5292 (Japan); Nobutoki, M. [Nippon Sanso Co., Kawasaki, Kanagawa, 210-0861 (Japan)

2004-06-23T23:59:59.000Z

95

Helium release from radioisotope heat sources  

DOE Green Energy (OSTI)

Diffusion of helium in /sup 238/PuO/sub 2/ fuel was characterized as a function of the heating rate and the fuel microstructure. The samples were thermally ramped in an induction furnace and the helium release rates measured with an automated mass spectrometer. The diffusion constants and activation energies were obtained from the data using a simple diffusion model. The release rates of helium were correlated with the fuel microstructure by metallographic examination of fuel samples. The release mechanism consists of four regimes, which are dependent upon the temperature. Initially, the release is controlled by movement of point defects combined with trapping along grain boundaries. This regime is followed by a process dominated by formation and growth of helium bubbles along grain boundaries. The third regime involves volume diffusion controlled by movement of oxygen vacancies. Finally, the release at the highest temperatures follows the diffusion rate of intragranular bubbles. The tendency for helium to be trapped within the grain boundaries diminishes with small grain sizes, slow thermal pulses, and older fuel.

Peterson, D.E.; Early, J.W.; Starzynski, J.S.; Land, C.C.

1984-05-01T23:59:59.000Z

96

Helium-filled aluminum flight tubes  

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

Helium-filled aluminum flight tubes. Helium-filled aluminum flight tubes. Detector housing for the CCD camera lens, mirror, and scintillator. For more information, contact Instrument Scientist: Hassina Bilheux, bilheuxhn@ornl.gov, 865.384.9630 neutrons.ornl.gov/instruments/HFIR/factsheets/Instrument-cg1d.pdf The CG-1D beam is used for neutron imaging measurements using a white beam. Apertures (with different diameters D (pinhole geometry) are used at the entrance of the helium-filled flight path to allow L/D variation from 400 to 800. L is the distance between the aperture and the detector (where the image is produced). Samples sit on a translation/ rotation stage for alignment and tomography purposes. Detectors for CG-1D include

97

SCREW COMPRESSOR CHARACTERISTICS FOR HELIUM REFRIGERATION SYSTEMS  

SciTech Connect

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

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

2008-03-01T23:59:59.000Z

98

GAS COOLED POWER REACTOR COOLANT CHOICE  

SciTech Connect

The current status of helium and carbon dioxide technology is described in the light of the Gas Cooled Reactor Program requiremoents. The problem of containing high-pressure helium at high temperature is discussed, and it is concluded that, by proper attention to the design, construction and maintenance of a plant, a high degree of helium leak-tightness can be achieved at small additional cost when compared with a carbon dioxide system. What is more, the cost of making up helium losses in a practically achievable system is estimated to be small compared with other fixed and operating costs. Graphite-carbon dioxide reaction data are reviewed. It is shown that carbon dioxide at atmospheric pressure and low flow rates should be compatible with a graphite mooderator up to 525 C. No data are available at the high pressures and fiow rates that would be encountered in power reactors. Significantiy higher oxidation rates may result, however, perhaps limiting bulk moderator temperatures to 450 to 500 C. Improved carbon materials, protective coatings and inhibitors, and/or operating practices may be developed that will allow significant future increases in these limiting temperatures. (auth)

Heacock, H.W.; Nightingale, R.E.

1958-06-12T23:59:59.000Z

99

Properties of hydrogen/helium accretion plasmas  

DOE Green Energy (OSTI)

We study the properties of impulsively-heated plasmas initially composed of hydrogen and helium. We follow the time-dependent behavior of the ion and electron temperatures, the pair density, and the densities of hydrogen, helium, and nuclei formed in fusion and breakup reactions. We also consider neutron production and escape, and calculate the 0.431 and 0.478 MeV line luminosities from ..cap alpha..-..cap alpha.. fusion reactions, and the 2.22 MeV line luminosity from neutron capture on protons. 13 refs., 3 figs., 1 tab.

Guessoum, N.; Dermer, C.D.

1988-02-01T23:59:59.000Z

100

Lessons Learned From Gen I Carbon Dioxide Cooled Reactors  

Science Conference Proceedings (OSTI)

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.

David E. Shropshire

2004-04-01T23:59:59.000Z

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101

STUDY OF HELIUM RETENTION IN NANO-CAVITY ...  

Science Conference Proceedings (OSTI)

STUDY OF HELIUM RETENTION IN NANO-CAVITY TUNGSTEN AS A FIRST WALL IN A FUSION CHAMBER USING NEUTRON DEPTH ...

102

Search for the eta-mesic helium at COSY  

E-Print Network (OSTI)

We review status and perspectives of the search of the eta-mesic helium at the cooler synchrotron COSY.

P. Moskal

2009-09-22T23:59:59.000Z

103

Design Windows for a He Cooled Fusion Reactor* Dai-Kai Sze and Ahmed Hassanein  

E-Print Network (OSTI)

Design Windows for a He Cooled Fusion Reactor* Dai-Kai Sze and Ahmed Hassanein Argonne National Laboratory 9700 South Cass Avenue, Argonne, IL 60439 EQUATIONDERIVATION ABSTRACT A design window concept by this design window concept. INTRODUCTION Helium is an attractive coolant for both fusion and fission reactors

Harilal, S. S.

104

Estimation of radiation damage at the structural materials of a hybrid reactor by probabilistic neural networks  

Science Conference Proceedings (OSTI)

This paper presents a new approach based on probabilistic neural networks (PNNs) for the radiation damage parameters at the structural material of a nuclear fusion-fission (hybrid) reactor. Artificial neural networks (ANNs) have recently been introduced ... Keywords: Atomic displacement, Helium generation, Hybrid reactor, Probabilistic neural networks (PNNs), Radiation damage

Elif Derya íbeyli; Mustafa íbeyli

2009-04-01T23:59:59.000Z

105

Fusion reactor high vacuum pumping: Charcoal cryosorber tritium exposure results  

SciTech Connect

Recent experiments, have shown the practically of using activated charcoal (coconut charcoal) at 4{degrees}K to pump helium and hydrogen isotopes for a fusion reactor. Both speed and capacity for deuterium/helium and tritium/helium-3 mixtures were shown to be satisfactory. The long term effects of tritium on the charcoal/cement system developed by Grumman and LLNL were not known and a program was undertaken to see what, if any, effect long term tritium exposure has on the cryosorber. Several charcoal on aluminum test samples were subjected to six months exposure of tritium at approximately 77{degrees}K. The tritium was scanned several times with a residual gas analyzer and the speed-capacity performance of the samples was measured before, approximately half way through and after the exposure. Modest effects were noted which would not seriously restrict charcoal's use as a cryosorber for fusion reactor high vacuum pumping applications. 4 refs., 8 figs.

Sedgley, D.W.; Walthers, C.R.; Jenkins, E.M. (Grumman Aerospace Corp., Bethpage, NY (United States))

1991-01-01T23:59:59.000Z

106

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

SciTech Connect

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.

Koch, M.; Kazimi, M.S.

1991-04-01T23:59:59.000Z

107

HEAVY WATER MODERATED NEUTRONIC REACTOR  

DOE Patents (OSTI)

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.

Szilard, L.

1958-04-29T23:59:59.000Z

108

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Treshow, M.

1961-09-01T23:59:59.000Z

109

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

Daniels, F.

1959-10-27T23:59:59.000Z

110

Operating experience with gas-bearing circulators in a high-pressure helium loop  

Science Conference Proceedings (OSTI)

A high-pressure engineering test loop has been designed and constructed at the Oak Ridge National Laboratory for circulating helium through a test chamber at temperatures to 1000/sup 0/C. The purpose of this loop is to determine the thermal and structural performance of proposed components for the primary loops of gas-cooled nuclear reactors. Five MW of power is available to provide the required gas temperature at the test chamber, and an air-cooled heat exchanger, rated at 4.4 MW, serves as a heat sink. This report contains results of tests performed on gas-bearing circulators.

Sanders, J.P.; Gat, Uri; Young, H.C.

1987-01-01T23:59:59.000Z

111

CONVECTION REACTOR  

DOE Patents (OSTI)

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

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

1960-03-22T23:59:59.000Z

112

lullIIlllIllLLLII DESIGN WINDOWS FOR A He COOLED FUSION REACTOR*  

E-Print Network (OSTI)

_ ii|l iImMmmm lullIIlllIllLLLII #12;. #12;DESIGN WINDOWS FOR A He COOLED FUSION REACTOR '....."[',":-,_.30 B93 _I_TFIII_31_ONOF THIS DO_.JMENT IS LJNLIMITED 0 S 1" I #12;,l° Design Windows for a He Cooled A design window concept is developed for a He-cooled of a helium cooled reactor are: fusion reactor blanket

Harilal, S. S.

113

Time dependence of liquid-helium fluorescence  

SciTech Connect

The time dependence of extreme ultraviolet (EUV) fluorescence following an ionizing radiation event in liquid helium is observed and studied in the temperature range from 250 mK to 1.8 K. The fluorescence exhibits significant structure including a short ({approx}10 ns) strong initial pulse followed by single photons whose emission rate decays exponentially with a 1.6-{mu}s time constant. At an even longer time scale, the emission rate varies as '1/time' (inversely proportional to the time after the initial pulse). The intensity of the '1/time' component from {beta} particles is significantly weaker than those from {alpha} particles or neutron capture on {sup 3}He. It is also found that for {alpha} particles, the intensity of this component depends on the temperature of the superfluid helium. Proposed models describing the observed fluorescence are discussed.

McKinsey, D.N.; Brome, C.R.; Dzhosyuk, S.N.; Mattoni, C.E.H.; Yang, L.; Doyle, J.M. [Department of Physics, Harvard University, Cambridge, Massachusetts 02138 (United States); Golub, R.; Habicht, K.; Korobkina, E. [Hahn-Meitner Institut, Berlin-Wannsee (Germany); Huffman, P.R.; Thompson, A.K. [National Institute of Standards and Technology, Gaithersburg, Maryland 20899 (United States); Lamoreaux, S.K. [Los Alamos National Laboratory, Los Alamos, New Mexico 87544 (United States)

2003-06-01T23:59:59.000Z

114

Factors influencing helium measurements for detection of control rod failures in BWR  

SciTech Connect

Much effort has been made to minimize the number and consequences of fuel failures at nuclear power plants. The consequences of control rod failures have also gained an increased attention. In this paper we introduce a system for on-line surveillance of control rod integrity which has several advantages comparing to the surveillance methods available today in boiling water reactors (BWRs). This system measures the helium released from failed control rods containing boron carbide (B4C). However, there are a number of factors that might influence measurements, which have to be taken into consideration when evaluating the measured data. These factors can be separated into two groups: 1) local adjustments, made on the sampling line connecting the detector to the off-gas system, and 2) plant operational parameters. The adjustments of the sample line conditions include variation of gas flow rate and gas pressure in the line. Plant operational factors that may influence helium measurements can vary from plant to plant. The factors studied at Leibstadt nuclear power plant (KKL) were helium impurities in injected hydrogen gas, variation of the total off-gas flow and regular water refill. In this paper we discuss these factors and their significance and present experimental results of measurements at KKL. (authors)

Larsson, I.; Sihver, L. [Div. of Nuclear Engineering, Dept. of Applied Physics, Chalmers Univ. of Technology, SE-412 96 Gothenburg (Sweden); Loner, H.; Ledergerber, G. [Kernkraftwerk Leibstadt, CH-5325 Leibstadt (Switzerland); Schnurr, B. [E.ON Kernkraft GmbH, D-84049 Essenbach (Germany)

2012-07-01T23:59:59.000Z

115

Combined cold compressor/ejector helium refrigerator  

DOE Patents (OSTI)

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.

Brown, Donald P. (Southold, NY)

1985-01-01T23:59:59.000Z

116

Combined cold compressor/ejector helium refrigerator  

DOE Patents (OSTI)

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.

Brown, D.P.

1984-06-05T23:59:59.000Z

117

Studies of Helium Distribution in Metal Tritides  

SciTech Connect

The distribution of helium (3HE) in LiT, TiT2, and UT3, which are regarded as representative metal tritides, was investigated using pulse nuclear magnetic resonance (NMR) techniques. Analyses of the NMR lineshapes and nuclear relaxation times indicate the 3He atoms are trapped in microscopic gas bubbles for each tritide. The effects of concentration and temperature on the 3He distributions were investigated as well.

Bowman, Jr., R. C.; Attalla, A.

1976-03-01T23:59:59.000Z

118

The Helium spectrum in erupting solar prominences  

E-Print Network (OSTI)

Even quiescent solar prominences may become active and sometimes erupt. These events are occasionally linked to coronal mass ejections. However we know very little about the plasma properties during the activation and eruption processes. We present new computations of the helium line profiles emitted by an eruptive prominence. The prominence is modelled as a plane-parallel slab standing vertically above the solar surface and moving upward as a solid body. The helium spectrum is computed with a non local thermodynamic equilibrium radiative transfer code. The effect of Doppler dimming / brightening is investigated in the resonance lines of He I and He II formed in the EUV, as well as on the He I 10830 A and 5876 A lines. We focus on the line profile properties and the resulting integrated intensities. It is shown that the helium lines are very sensitive to Doppler dimming effects. We also study the effect of frequency redistribution in the formation mechanisms of the resonance lines and find that it is necessary to use partial redistribution in frequency for the resonance lines.

Nicolas Labrosse; Pierre Gouttebroze; Jean-Claude Vial

2006-09-18T23:59:59.000Z

119

A Helium Isotope Perspective On The Dixie Valley, Nevada, Hydrothermal  

Open Energy Info (EERE)

Helium Isotope Perspective On The Dixie Valley, Nevada, Hydrothermal Helium Isotope Perspective On The Dixie Valley, Nevada, Hydrothermal System Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: A Helium Isotope Perspective On The Dixie Valley, Nevada, Hydrothermal System Details Activities (3) Areas (1) Regions (0) Abstract: Fluids from springs, fumaroles, and wells throughout Dixie Valley, NV were analyzed for noble gas abundances and isotopic compositions. The helium isotopic compositions of fluids produced from the Dixie Valley geothermal field range from 0.70 to 0.76 Ra, are among the highest values in the valley, and indicate that similar to 7.5% of the total helium is derived from the mantle. A lack of recent volcanics or other potential sources requires flow of mantle-derived helium up along the

120

Trace Detection of Metastable Helium Molecules in Superfluid Helium by Laser-Induced Fluorescence  

SciTech Connect

We describe an approach to detecting ionizing radiation that combines the special properties of superfluid helium with the sensitivity of quantum optics techniques. Ionization in liquid helium results in the copious production of metastable He{sub 2} molecules, which can be detected by laser-induced fluorescence. Each molecule can be probed many times using a cycling transition, resulting in the detection of individual molecules with high signal to noise. This technique could be used to detect neutrinos, weakly interacting massive particles, and ultracold neutrons, and to image superfluid flow in liquid {sup 4}He.

McKinsey, D. N.; Lippincott, W.H.; Nikkel, J.A.; Rellergert, W.G. [Department of Physics, Yale University, P.O. Box 208120, New Haven, Connecticut 06520 (United States)

2005-09-09T23:59:59.000Z

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


121

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

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

1961-11-21T23:59:59.000Z

122

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

SciTech Connect

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 information, and performance of the alloys in helium containing a wide range of impurity concentrations. A detailed thermal hydraulic analysis, using a model developed at ANL, was performed to calculate heat transfer, temperature distribution, and pressure drop inside both printed circuit and shell-and-tube heat exchangers. The analysis included evaluation of the role of key process parameters, geometrical factors in HX designs, and material properties. Calculations were performed for helium-to-helium, helium-to-helium/nitrogen, and helium-to-salt HXs. The IHX being a high temperature component, probably needs to be designed using ASME Code Section III, Subsection NH, assuming that the IHX will be classified as a class 1 component. With input from thermal hydraulic calculations performed at ANL, thermal conduction and stress analyses for both compact and shell-and-tube HXs were performed.

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

2007-04-05T23:59:59.000Z

123

REACTOR COOLING  

DOE Patents (OSTI)

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

Quackenbush, C.F.

1959-09-29T23:59:59.000Z

124

The Potential of Helium as a Guide to Uranium Ore  

Science Conference Proceedings (OSTI)

Martin Marietta Corp. and Earth Sciences, Inc., studied the effectiveness of helium surveying as a tool for uranium exploration. They generated basic data on the little-known distribution of helium in soils, tested various techniques for conducting surveys in the field, developed guidelines for helium surveys and interpretation, and stimulated interest, if it was warranted, in the further testing and application of promising approaches.

1978-07-01T23:59:59.000Z

125

Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1980-March 31, 1980  

Science Conference Proceedings (OSTI)

Results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Included are the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described, including screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, and 950/sup 0/C.

Not Available

1980-06-25T23:59:59.000Z

126

The Helium-3 Supply Crisis – Alternative Techniques to ...  

Science Conference Proceedings (OSTI)

... Helium-3 is a by-product of tritium decay after 12 years. Due to the nonetheless encouraging reduction in stocks of tritium ...

2010-04-16T23:59:59.000Z

127

NIST Studies How New Helium Ion Microscope Measures Up  

Science Conference Proceedings (OSTI)

... are studying helium ion microscopes to improve ... analogous to the scanning electron microscope, which was ... are far larger than electrons, they can ...

2012-11-02T23:59:59.000Z

128

Mantle Helium And Carbon Isotopes In Separation Creek Geothermal...  

Open Energy Info (EERE)

Mantle Helium And Carbon Isotopes In Separation Creek Geothermal Springs, Three Sisters Area, Central Oregon- Evidence For Renewed Volcanic Activity Or A Long Term Steady State...

129

Regional And Local Trends In Helium Isotopes, Basin And Range...  

Open Energy Info (EERE)

Superimposed on this general regional trend are isolated features with elevated helium isotope ratios (0.8-2.1 Ra) compared to the local background. Spring geochemistry and...

130

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

Wigner, E.P.

1958-04-22T23:59:59.000Z

131

Simulation program for central helium liquefier  

SciTech Connect

The computer program described here analyzes the performance of Fermilab Central Helium Liquefier (CHL) and predicts the values of the plant thermodynamic variables at all process points in the plant. To simulate CHL, this program is modified from the prototype program which was developed by Hitachi Ltd. a couple of years ago. This program takes care of only the steady state simulation and takes account of the change of the turbine efficiency, the pressure drops and the UA values of the heat exchangers. How to use the program is shown.

Kawamura, S.

1984-02-20T23:59:59.000Z

132

DIRECT EVALUATION OF THE HELIUM ABUNDANCES IN OMEGA CENTAURI  

SciTech Connect

A direct measure of the helium abundances from the near-infrared transition of He I at 1.08 {mu}m is obtained for two nearly identical red giant stars in the globular cluster Omega Centauri. One star exhibits the He I line; the line is weak or absent in the other star. Detailed non-local thermal equilibrium semi-empirical models including expansion in spherical geometry are developed to match the chromospheric H{alpha}, H{beta}, and Ca II K lines, in order to predict the helium profile and derive a helium abundance. The red giant spectra suggest a helium abundance of Y {<=} 0.22 (LEID 54064) and Y = 0.39-0.44 (LEID 54084) corresponding to a difference in the abundance {Delta}Y {>=} 0.17. Helium is enhanced in the giant star (LEID 54084) that also contains enhanced aluminum and magnesium. This direct evaluation of the helium abundances gives observational support to the theoretical conjecture that multiple populations harbor enhanced helium in addition to light elements that are products of high-temperature hydrogen burning. We demonstrate that the 1.08 {mu}m He I line can yield a helium abundance in cool stars when constraints on the semi-empirical chromospheric model are provided by other spectroscopic features.

Dupree, A. K.; Avrett, E. H., E-mail: dupree@cfa.harvard.edu, E-mail: eavrett@cfa.harvard.edu [Harvard-Smithsonian Center for Astrophysics, Cambridge, MA 02138 (United States)

2013-08-20T23:59:59.000Z

133

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

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

1958-12-01T23:59:59.000Z

134

Decommissioning the UHTREX Reactor Facility at Los Alamos, New Mexico  

SciTech Connect

The Ultra-High Temperature Reactor Experiment (UHTREX) facility was constructed in the late 1960s to advance high-temperature and gas-cooled reactor technology. The 3-MW reactor was graphite moderated and helium cooled and used 93% enriched uranium as its fuel. The reactor was run for approximately one year and was shut down in February 1970. The decommissioning of the facility involved removing the reactor and its associated components. This document details planning for the decommissioning operations which included characterizing the facility, estimating the costs of decommissioning, preparing environmental documentation, establishing a system to track costs and work progress, and preplanning to correct health and safety concerns in the facility. Work to decommission the facility began in 1988 and was completed in September 1990 at a cost of $2.9 million. The facility was released to Department of Energy for other uses in its Los Alamos program.

Salazar, M.; Elder, J.

1992-08-01T23:59:59.000Z

135

Design Considerations for Economically Competitive Sodium Cooled Fast Reactors  

SciTech Connect

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.

Hongbin Zhang; Haihua Zhao

2009-05-01T23:59:59.000Z

136

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

SciTech Connect

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.

Contescu, Cristian I [ORNL] ORNL

2013-09-01T23:59:59.000Z

137

Reactor Materials  

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

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

138

NEUTRONIC REACTORS  

DOE Patents (OSTI)

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

Wigner, E.P.

1960-11-22T23:59:59.000Z

139

REACTOR SHIELD  

DOE Patents (OSTI)

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

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

1959-02-17T23:59:59.000Z

140

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

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

1962-10-23T23:59:59.000Z

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


141

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

DOE Patents (OSTI)

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.

Dean, John W. (Los Alamos, NM)

1977-01-01T23:59:59.000Z

142

Testing CPT Invariance with Antiprotonic Helium Atoms  

Science Conference Proceedings (OSTI)

The structure of matter is related to symmetries at every level of study. CPT symmetry is one of the most important laws of field theory: it states the invariance of physical properties when one simultaneously changes the signs of the charge and of the spatial and time coordinates of free elementary particles. Although in general opinion CPT symmetry is not violated in Nature, there are theoretical attempts to develop CPT-violating models. The Antiproton Decelerator at CERN has been built to test CPT invariance. The ASACUSA experiment compares the properties of particles and antiparticles by studying the antiprotonic helium atom via laser spectroscopy and measuring the mass, charge and magnetic moment of the antiproton as compared to those of the proton.

Horvath, Dezso [KFKI Research Institute for Particle and Nuclear Physics, H-1525 Budapest, Hungary and Institute of Nuclear Research (ATOMKI), Debrecen (Hungary)

2008-08-08T23:59:59.000Z

143

Compact hydrogen/helium isotope mass spectrometer  

DOE Patents (OSTI)

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).

Funsten, Herbert O. (Los Alamos, NM); McComas, David J. (Los Alamos, NM); Scime, Earl E. (Morgantown, WV)

1996-01-01T23:59:59.000Z

144

THE EXPERIMENTAL BERYLLIUM OXIDE REACTOR. MARITIME GAS-COOLED REACTOR PROGRAM  

SciTech Connect

LIUM OXIDE REACTOR. MARITIME GAS-COOLED The Experimental Beryllium Oxide Reactor, EBOR, will be constructed at the National Reactor Testing Station as the AEC portion of the joint Maritime Administration--AEC Maritime Gas Cooled Reactor Program. The ultimate goal of the Program is the development of nuclear power plants employing a helium cooled and beryllium oxide moderated reactor directly coupled to a closed cycle gas turbine. The objective is to obtain compact nuclear engines suitable for use either in a merchant ship propulsion system or an intermediate size central station power plant in the 20 to 100 Mw(e) size range. The EBOR is a l0 Mw(t) test of the basic fuel element and moderator designs. It is capable of being up-graded in power at a later date to a test of the nuclear reactor turbine concept. The objective of the experiment is outlined. The principal reactor components to be tested and the test facility are described. (auth)

Moore, W.C.

1961-07-01T23:59:59.000Z

145

ACOUSTIC SIGNATURES OF THE HELIUM CORE FLASH  

SciTech Connect

All evolved stars with masses M {approx}< 2 M{sub Sun} undergo an initiating off-center helium core flash in their M{sub c} Almost-Equal-To 0.48 M{sub Sun} He core as they ascend the red giant branch (RGB). This off-center flash is the first of a few successive helium shell subflashes that remove the core electron degeneracy over 2 Myr, converting the object into a He-burning star. Though characterized by Thomas over 40 years ago, this core flash phase has yet to be observationally probed. Using the Modules for Experiments in Stellar Astrophysics (MESA) code, we show that red giant asteroseismology enabled by space-based photometry (i.e., Kepler and CoRoT) can probe these stars during the flash. The rapid ({approx}< 10{sup 5} yr) contraction of the red giant envelope after the initiating flash dramatically improves the coupling of the p-modes to the core g-modes, making the detection of l = 1 mixed modes possible for these 2 Myr. This duration implies that 1 in 35 stars near the red clump in the H-R diagram will be in their core flash phase. During this time, the star has a g-mode period spacing of {Delta}P{sub g} Almost-Equal-To 70-100 s, lower than the {Delta}P{sub g} Almost-Equal-To 250 s of He-burning stars in the red clump, but higher than the RGB stars at the same luminosity. This places them in an underpopulated part of the large frequency spacing ({Delta}{nu}) versus {Delta}P{sub g} diagram that should ease their identification among the thousands of observed red giants.

Bildsten, Lars; Paxton, Bill; Moore, Kevin; Macias, Phillip J. [Kavli Institute for Theoretical Physics and Department of Physics Kohn Hall, University of California, Santa Barbara, CA 93106 (United States)

2012-01-15T23:59:59.000Z

146

Research reactors - an overview  

SciTech Connect

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

West, C.D.

1997-03-01T23:59:59.000Z

147

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

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

1958-09-01T23:59:59.000Z

148

Brookhaven Medical Research Reactor  

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

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

149

Evolution of the Standard Helium Liquefier and Refrigerator Range designed by Air Liquide DTA, France  

E-Print Network (OSTI)

Evolution of the Standard Helium Liquefier and Refrigerator Range designed by Air Liquide DTA, France

Crispel, S; Caillaud, A; Delcayre, F; Grabie, V

2008-01-01T23:59:59.000Z

150

Transient Behaviour and Helium Discharge in Cryogenic Distribution Line (QRL) Headers Following Breakdown of Insulation Vacuum  

E-Print Network (OSTI)

Transient Behaviour and Helium Discharge in Cryogenic Distribution Line (QRL) Headers Following Breakdown of Insulation Vacuum

Chorowski, M

1997-01-01T23:59:59.000Z

151

CALIOP: a multichannel design code for gas-cooled fast reactors. Code description and user's guide  

Science Conference Proceedings (OSTI)

CALIOP is a design code for fluid-cooled reactors composed of parallel fuel tubes in hexagonal or cylindrical ducts. It may be used with gaseous or liquid coolants. It has been used chiefly for design of a helium-cooled fast breeder reactor and has built-in cross section information to permit calculations of fuel loading, breeding ratio, and doubling time. Optional cross-section input allows the code to be used with moderated cores and with other fuels.

Thompson, W.I.

1980-10-01T23:59:59.000Z

152

REACTOR CONTROL  

DOE Patents (OSTI)

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

Fortescue, P.; Nicoll, D.

1962-04-24T23:59:59.000Z

153

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Christy, R.F.

1958-07-15T23:59:59.000Z

154

POWER REACTOR  

DOE Patents (OSTI)

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

Zinn, W.H.

1958-07-01T23:59:59.000Z

155

Catalytic reactor  

DOE Patents (OSTI)

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

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

2009-03-10T23:59:59.000Z

156

NEUTRONIC REACTORS  

DOE Patents (OSTI)

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

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

1958-10-14T23:59:59.000Z

157

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Young, G.

1963-01-01T23:59:59.000Z

158

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

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

1958-04-15T23:59:59.000Z

159

Power Burst Facility (PBF) Reactor Reactor Decommissioning  

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

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

160

Mantle Helium And Carbon Isotopes In Separation Creek Geothermal Springs,  

Open Energy Info (EERE)

Mantle Helium And Carbon Isotopes In Separation Creek Geothermal Springs, Mantle Helium And Carbon Isotopes In Separation Creek Geothermal Springs, Three Sisters Area, Central Oregon- Evidence For Renewed Volcanic Activity Or A Long Term Steady State System(Question) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Conference Paper: Mantle Helium And Carbon Isotopes In Separation Creek Geothermal Springs, Three Sisters Area, Central Oregon- Evidence For Renewed Volcanic Activity Or A Long Term Steady State System(Question) Details Activities (1) Areas (1) Regions (0) Abstract: Here we present the helium and carbon isotope results from the initial study of a fluid chemistry-monitoring program started in the summer of 2001 near the South Sister volcano in central Oregon. The Separation Creek area which is several miles due west of the volcano is the locus of

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


161

Shock Experiments on Pre-Compressed Fluid Helium  

Science Conference Proceedings (OSTI)

We summarize current methods and results for coupling laser-induced shocks into pre-compressed Helium contained in a diamond anvil cell (DAC). We are able to load helium, hydrogen, deuterium, and helium-hydrogen mixtures into a DAC and propagate a laser-generated shock into the pre-compressed sample. This technique has allowed us to measure the Hugoniot for helium at initial densities ranging from 1 to 3.5 times liquid density. We have developed and used a methodology whereby all of our measurements are referenced to crystalline quartz, which allows us to update our results as the properties of quartz are refined in the future. We also report the identification and elimination of severe electro-magnetic pulses (EMP) associated with plasma stagnation associated with ablation in a DAC.

Eggert, J. H.; Celliers, P. M.; Hicks, D. G.; Rygg, J. R.; Collins, G. W. [Lawrence Livermore National Laboratory, Livermore CA (United States); Brygoo, S.; Loubeyre, P. [Commissariat a l'Energie Atomique, Bruyeres-le-Chatel (France); McWilliams, R. S.; Spaulding, D.; Jeanloz, R. [University of California, Berkeley, CA (United States); Boehly, T. R. [University of Rochester, Rochester, NY (United States)

2009-09-10T23:59:59.000Z

162

The Liquefaction of Hydrogen and Helium Using Small Coolers  

E-Print Network (OSTI)

because the J-T valve and heat exchanger clogged because ofseparate J-T valve and J-T heat exchanger had to be used. Itadded J-T circuit and heat exchanger, to liquefy helium have

Green, Michael A.

2006-01-01T23:59:59.000Z

163

SRS upgrades helium recovery system | National Nuclear Security  

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

upgrades helium recovery system | National Nuclear Security upgrades helium recovery system | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > NNSA Blog > SRS upgrades helium recovery system SRS upgrades helium recovery system Posted By Office of Public Affairs Savannah River Site (SRS) Tritium Programs recently completed a project to design, build and relocate a new system for separating and capturing

164

L3 Milestone Use Computational Model to Design and Optimize Welding Conditions to Suppress Helium  

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

L3 Milestone L3 Milestone Use Computational Model to Design and Optimize Welding Conditions to Suppress Helium Cracking during Welding June 2012 Wei Zhang and Zhili Feng, ORNL Eric Willis, EPRI Background and Objectives Today, welding is widely used for repair, maintenance and upgrade of nuclear reactor components. As a critical technology to extend the service life of nuclear power plants beyond 60 years, weld technology must be further developed to meet new challenges associated with the aging of the plants, such as control and mitigation of the detrimental effects of weld residual stresses and repair of highly irradiated materials. To meet this goal, fundamental understanding of the "welding" effect is necessary for development of new and improved welding technologies.

165

Operations aspects of the Fermilab Central Helium Liquefier Facility  

SciTech Connect

The Fermilab Central Helium Liquefier (CHL) facility consists of helium and nitrogen reliquefier plants operated 24 hours-a-day to supply LHe at 4.6{degrees}K and LN{sub 2} for the Fermilab Tevatron superconducting proton-antiproton collider ring and to recover warm return gases. Operating aspects of CHL, including different equipment and systems reliability, availability, maintenance experience, safety concerns, and economics aspects are discussed.

Geynisman, M.G.; Makara, J.N.

1995-03-01T23:59:59.000Z

166

Operations aspects of the Fermilab Central Helium Liquefier facility  

SciTech Connect

The Fermilab Central Helium Liquefier (CHL) facility consists of helium and nitrogen reliquefier plants operated 24 hours-a-day to supply LHe at 4.6 K and LN{sub 2} for the Fermilab Tevatron superconducting proton-antiproton collider ring and to recover warm return gases. Operating aspects of CHL, including different equipment and systems reliability, availability, maintenance experience, safety concerns, and economics aspects are discussed.

Geynisman, M.G.; Makara, J.N.

1996-09-01T23:59:59.000Z

167

Recovery of purified helium or hydrogen from gas mixtures  

DOE Patents (OSTI)

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)

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

1974-01-15T23:59:59.000Z

168

NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

Wigner, E.P.

1962-12-25T23:59:59.000Z

169

Neutronic reactor  

DOE Patents (OSTI)

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

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

1976-08-17T23:59:59.000Z

170

NEUTRONIC REACTORS  

DOE Patents (OSTI)

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

Anderson, H.L.

1958-10-01T23:59:59.000Z

171

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Post, R.G.

1963-05-01T23:59:59.000Z

172

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Starr, C.

1963-01-01T23:59:59.000Z

173

REACTOR UNLOADING  

DOE Patents (OSTI)

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

Leverett, M.C.

1958-02-18T23:59:59.000Z

174

NUCLEAR REACTOR  

DOE Patents (OSTI)

A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

Treshow, M.

1958-08-19T23:59:59.000Z

175

Neutronic reactor  

DOE Patents (OSTI)

A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

Lewis, Warren R. (Richland, WA)

1978-05-30T23:59:59.000Z

176

NUCLEAR REACTORS  

DOE Patents (OSTI)

An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

1961-12-01T23:59:59.000Z

177

REACTOR CONTROL  

DOE Patents (OSTI)

This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

Ruano, W.J.

1957-12-10T23:59:59.000Z

178

Light Water Reactors Technology Development - Nuclear Reactors  

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

Light Water Reactors Light Water Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

179

Containment Versus Confinement for High-Temperature Gas Reactors: Regulatory, Design Basis, Siting, and Cost/Economic Considerations  

Science Conference Proceedings (OSTI)

This report provides the results of an investigation pertaining to the use of the confinement that has been proposed for the high temperature and very high temperature gas reactors (HTGR, VHTR). No comprehensive study of this question has been published since 1985. All power reactor designs to go into commercial service in the United States were light water reactors (LWR), except for Fort St. Vrain (FSV) and Peach Bottom Unit 1, which were steam cycle helium gas cooled reactors. All designs use a leak-ti...

2005-05-04T23:59:59.000Z

180

David Lee, Douglas Osheroff, Superfluidity, and Helium 3  

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

David Lee, Douglas Osheroff, Superfluidity, and Helium 3 David Lee, Douglas Osheroff, Superfluidity, and Helium 3 Resources with Additional Information David M. Lee and Douglas D. Osheroff received the 1996 Nobel Prize in Physics for 'their discovery of superfluidity in helium-3'. "In 1976, Lee shared with Richardson and Osheroff their earliest recognition for studies of superfluidity, the Simon Memorial Prize of the British Physical Society. The Buckley Prize of the American Physical Society followed for the trio in 1981. ... Douglas D. Osheroff Douglas D. Osheroff Courtesy AIP Emilio Segrè Visual Archives, AIP Meggers Gallery of Nobel Laureates David Lee David M. Lee Photo by Janerik Henriksson, Courtesy AIP Emilio Segrè Visual Archievs, W.F. Meggers Gallery of Nobel Laureate From 1966-67, Lee was a visiting scientist at Brookhaven National Laboratory ... ."1 His research at Cornell University includes " Low Temperature Physics: Normal and superfluid 3He, studies of the orientation of solid helium by optical birefringence, solid 3He and 4He, lambda phase diagram of 3He - 4He mixtures, quasiparticle tunneling in superconductors, magnetic resonance and ultrasound techniques, cooling by adiabatic demagnetization and the Pomeranchuk technique, spin waves in spin polarized hydrogen gas, high concentrations of hydrogen and nitrogen atons via matrix isolation by impurity-helium solids, magnetism, electron spin resonance."2

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


181

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

DOE Green Energy (OSTI)

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.

Lee O. Nelson

2011-04-01T23:59:59.000Z

182

Nuclear reactor  

DOE Patents (OSTI)

A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

Pennell, William E. (Greensburg, PA); Rowan, William J. (Monroeville, PA)

1977-01-01T23:59:59.000Z

183

ELECTRONUCLEAR REACTOR  

DOE Patents (OSTI)

An electronuclear reactor is described in which a very high-energy particle accelerator is employed with appropriate target structure to produce an artificially produced material in commercial quantities by nuclear transformations. The principal novelty resides in the combination of an accelerator with a target for converting the accelerator beam to copious quantities of low-energy neutrons for absorption in a lattice of fertile material and moderator. The fertile material of the lattice is converted by neutron absorption reactions to an artificially produced material, e.g., plutonium, where depleted uranium is utilized as the fertile material.

Lawrence, E.O.; McMillan, E.M.; Alvarez, L.W.

1960-04-19T23:59:59.000Z

184

Photocatalytic reactor  

DOE Patents (OSTI)

A photocatalytic reactor for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane.

Bischoff, Brian L. (Knoxville, TN); Fain, Douglas E. (Oak Ridge, TN); Stockdale, John A. D. (Knoxville, TN)

1999-01-01T23:59:59.000Z

185

Isotope Effects and Helium Retention Behavior in Vanadium Tritide  

SciTech Connect

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.

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

1985-04-01T23:59:59.000Z

186

CONTROL MEANS FOR REACTOR  

DOE Patents (OSTI)

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

Manley, J.H.

1961-06-27T23:59:59.000Z

187

An effective loading method of americium targets in fast reactors  

Science Conference Proceedings (OSTI)

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)

Ohki, Shigeo; Sato, Isamu; Mizuno, Tomoyasu; Hayashi, Hideyuki; Tanaka, Kenya [Japan Atomic Energy Agency, 4002, Narita-cho, O-arai-machi, Higashi-Ibaraki-gun, Ibaraki 311-1393 (Japan)

2007-07-01T23:59:59.000Z

188

Nuclear Reactor Accidents  

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

Reactor Accidents The accidents at the Three Mile Island (TMI) and Chernobyl nuclear reactors have triggered particularly intense concern about radiation hazards. The TMI accident,...

189

Principles of Reactor Physics  

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

Nuclear Reactor Physics M A Smith Argonne National Laboratory Nuclear Engineering Division Phone: 630-252-9747, Email: masmith@anl.gov Abstract: Nuclear reactor physics deals with...

190

Sensitivity Studies of Advanced Reactors Coupled to High Temperature Electrolysis (HTE) Hydrogen Production Processes  

DOE Green Energy (OSTI)

High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 °C to 950 °C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the steam or air sweep loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycle producing the highest efficiencies varied depending on the temperature range considered.

Edwin A. Harvego; Michael G. McKellar; James E. O'Brien; J. Stephen Herring

2007-04-01T23:59:59.000Z

191

A liquid helium cryogenic system design for the GEM magnet  

SciTech Connect

The Superconducting Super Collider (SSC) Gammas, Electrons, Muons (GEM) magnet is a large superconducting solenoid with a total mass of 1.05 {times} 10{sup 6} kg and a stored energy of 2.5 G. A cryogenic system to cool and to maintain the GEM magnet to liquid helium temperature is described. The system is designed to operate effectively under a variety of operating conditions, including cooldown/warm-up, steady state operations, and quench. Primary cooling during steady-state operation is based on natural circulation thermosiphon flow through cooling tubes in the solenoid support bobbin. Additional cooling loops are included for lead and joint cooling and conductor stabilization. A helium refrigerator/liquefier rated at 2 kill and 20 g/s will be specified to meet the refrigeration requirements. Cooldown of the magnet from 300 K to liquid nitrogen temperatures is accomplished using a counterflow helium-to-liquid-nitrogen heat exchanger independent of the helium refrigerator. The system incorporates provisions for maintenance access during accelerator beam operation.

Deis, G.; Warren, R.P. [Lawrence Livermore National Lab., CA (United States); Richied, D.E.; Martovetsky, N.N.; Krupczak, J.J.; Sidi-Yekhlef, A.; Pace, J.R.; Collins, C.A. [Superconducting Super Collider Lab., Dallas, TX (United States)

1993-06-01T23:59:59.000Z

192

Charge Distribution about an Ionizing Electron Track in Liquid Helium  

E-Print Network (OSTI)

The dependence on an applied electric field of the ionization current produced by an energetic electron stopped in liquid helium can be used to determine the spatial distribution of secondary electrons with respect to their geminate partners. An analytic expression relating the current and distribution is derived. The distribution is found to be non-Gaussian with a long tail at larger distances.

G. M. Seidel; T. M. Ito; A. Ghosh; B. Sethumadhavan

2013-10-04T23:59:59.000Z

193

The Liquefaction of Hydrogen and Helium Using Small Coolers  

DOE Green Energy (OSTI)

This report discusses the history of the liquefaction of hydrogen and helium using small coolers. This history dates form the 1960's when two stage GM coolers capable of reaching 7 K were used to liquefy helium and hydrogen by suing an added compressor and J-T circuit. Liquefaction using the added circuit failed to become mainstream because the J-T valve and heat exchanger clogged because of impurities in the gas being liquefied. Liquefaction using a GM cooler without an added J-T circuit proved to be difficult because the first stage was not used to pre-cool the gas coming to the second stage of the cooler. Once the gas being liquefied was pre-cooled using the cooler first stage, improvements in the liquefaction rates were noted. The advent of low temperature pulse tube cooler (down to 2.5 K) permitted one to achieve dramatic improvement is the liquefactions rates for helium. Similar but less dramatic improvements are expected for hydrogen as well. Using the PT-415 cooler, one can expect liquefaction rates of 15 to 20 liters per day for helium or hydrogen provided the heat leak into the cooler and the storage vessel is low. A hydrogen liquefier for MICE is presented at the end of this report.

Green, Michael A.

2006-02-12T23:59:59.000Z

194

TRANSMISSION ELECTRON MICROSCOPY STUDY OF HELIUM BEARING FUSION WELDS  

DOE Green Energy (OSTI)

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.

Tosten, M; Michael Morgan, M

2008-12-12T23:59:59.000Z

195

PULSATIONS IN HYDROGEN BURNING LOW-MASS HELIUM WHITE DWARFS  

SciTech Connect

Helium core white dwarfs (WDs) with mass M {approx}< 0.20 M {sub sun} undergo several Gyr of stable hydrogen burning as they evolve. We show that in a certain range of WD and hydrogen envelope masses, these WDs may exhibit g-mode pulsations similar to their passively cooling, more massive carbon/oxygen core counterparts, the ZZ Cetis. Our models with stably burning hydrogen envelopes on helium cores yield g-mode periods and period spacings longer than the canonical ZZ Cetis by nearly a factor of 2. We show that core composition and structure can be probed using seismology since the g-mode eigenfunctions predominantly reside in the helium core. Though we have not carried out a fully nonadiabatic stability analysis, the scaling of the thermal time in the convective zone with surface gravity highlights several low-mass helium WDs that should be observed in search of pulsations: NLTT 11748, SDSS J0822+2753, and the companion to PSR J1012+5307. Seismological studies of these He core WDs may prove especially fruitful, as their luminosity is related (via stable hydrogen burning) to the hydrogen envelope mass, which eliminates one model parameter.

Steinfadt, Justin D. R. [Department of Physics, Broida Hall, University of California, Santa Barbara, CA 93106 (United States); Bildsten, Lars [Kavli Institute for Theoretical Physics and Department of Physics, Kohn Hall, University of California, Santa Barbara, CA 93106 (United States); Arras, Phil, E-mail: jdrs@physics.ucsb.ed, E-mail: bildsten@kitp.ucsb.ed, E-mail: arras@virginia.ed [Department of Astronomy, University of Virginia, P.O. Box 400325, Charlottesville, VA 22904 (United States)

2010-07-20T23:59:59.000Z

196

Etching of Graphene Devices with a Helium Ion Beam  

E-Print Network (OSTI)

IM microscope image in Figure 2a, was He ion etched by sequen- tial imaging in high resolution. The grapheneEtching of Graphene Devices with a Helium Ion Beam Max C. Lemme, David C. Bell,,§ James R. Williams as for pos- sible nanoelectronics applications.1 3 Many experiments in the field are targeted at graphene

Lukin, Mikhail

197

Implantation and Characterization of Helium in Nuclear Materials at ...  

Science Conference Proceedings (OSTI)

Symposium, Accelerated Materials Evaluation for Nuclear Application Utilizing Test Reactors, Ion Beam Facilities and Modeling. Presentation Title, Implantation

198

Reactor and method of operation  

DOE Patents (OSTI)

A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

Wheeler, John A. (Princeton, NJ)

1976-08-10T23:59:59.000Z

199

Effect of helium injection on diffusion dominated air ingress accidents in pebble bed reactors  

E-Print Network (OSTI)

The primary objective of this thesis was to validate the sustained counter air diffusion (SCAD) method at preventing natural circulation onset in diffusion dominated air ingress accidents. The analysis presented in this ...

Yurko, Joseph Paul

2010-01-01T23:59:59.000Z

200

Design and Testing of Helium Cooled Components for DEMO and Synergies with Gen4 Reactor  

Science Conference Proceedings (OSTI)

Power Plants, Demo, and Next Steps / Eighteenth Topical Meeting on the Technology of Fusion Energy (Part 2)

Thomas Ihli

Note: This page contains sample records for the topic "helium reactor gt-mhr" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

PROPOSED HELIUM PURIFICATION SYSTEM FOR THE EXPERIMENTAL GAS-COOLED REACTOR (EGCR)  

DOE Green Energy (OSTI)

Liquid and dry processes suituble for the purification of gases by the removal of CO/sub 2/, H/sub 2/O, CO, H/sub 2/, and hydrocarbons are discussed. Recommendations are given for specific processes io be included in a "dry" (no liquid absorbents or chemicals used) purification system for the hellum coolant of the EGCR The recommended processes include (1) a catalytic converter for the oxidation of CO, H/sub 2/, and hydrocarbons to CO/sub 2/ and H/sub 2/O, (2) cooler-condensors for the removal of the bulk of the R/sub 2/O, (3) silica gel adsorbers to complete the removal of H/sub 2/O, and (4) Linde Molecular Sieve adsorbers for the removal of CO/sub 2/. No provisions are included for the planned removal of radioactive gases or particulates. (auth)

Anderson, F.A.

1959-10-16T23:59:59.000Z

202

Manhattan Project: Final Reactor Design and X-10, 1942-1943  

Office of Scientific and Technical Information (OSTI)

Schematic of the X-10 Graphite Reactor, Oak Ridge FINAL REACTOR DESIGN AND X-10 Schematic of the X-10 Graphite Reactor, Oak Ridge FINAL REACTOR DESIGN AND X-10 (Met Lab and Oak Ridge [Clinton], 1942-1943) Events > The Plutonium Path to the Bomb, 1942-1944 Production Reactor (Pile) Design, 1942 DuPont and Hanford, 1942 CP-1 Goes Critical, December 2, 1942 Seaborg and Plutonium Chemistry, 1942-1944 Final Reactor Design and X-10, 1942-1943 Hanford Becomes Operational, 1943-1944 Before any plutonium could be chemically separated from uranium for a bomb, however, that uranium would first have to be irradiated in a production pile. CP-1 had been a success as a scientific experiment, but the pile was built on such a small scale that recovering any significant amounts of plutonium from it was impractical. In the fall of 1942, scientists of the Met Lab had decided to build a second Fermi pile at Argonne as soon as his experiments on the first were completed and to proceed with the "Mae West" design for a helium-cooled production pile as well. When DuPont engineers assessed the Met Lab's plans in the late fall, they agreed that helium should be given first priority. They placed heavy water second and urged an all-out effort to produce more of this highly effective moderator. Bismuth and water were ranked third and fourth in DuPont's analysis. Priorities began to change when Enrico Fermi's CP-1 calculations demonstrated a higher value for the neutron reproduction factor k (for a theoretical reactor of infinite size) than anyone had anticipated. Met Lab scientists concluded that a water-cooled pile was now feasible. Crawford Greenewalt, head of the DuPont effort, continued, however, to support helium cooling.

203

Applications for reactor-pumped lasers  

Science Conference Proceedings (OSTI)

Nuclear reactor-pumped lasers (RPLs) have been developed in the US by the Department of Energy for over two decades, with the primary research occurring at Sandia National Laboratories and Idaho National Engineering Laboratory. The US program has experimentally demonstrated reactor-pumped lasing in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1,271, 1,733, 1,792, 2,032, 2,630, 2,650, and 3,370 nm with intrinsic efficiency as high as 2.5%. The major strengths of a reactor-pumped laser are continuous high-power operation, modular construction, self-contained power, compact size, and a variety of wavelengths (from visible to infrared). These characteristics suggest numerous applications not easily accessible to other laser types. The continuous high power of an RPL opens many potential manufacturing applications such as deep-penetration welding and cutting of thick structures, wide-area hardening of metal surfaces by heat treatment or cladding application, wide-area vapor deposition of ceramics onto metal surfaces, production of sub-micron sized particles for manufacturing of ceramics, and 3-D ceramic lithography. In addition, a ground-based RPL could beam its power to space for such activities as illuminating geosynchronous communication satellites in the earth`s shadow to extend their lives, beaming power to orbital transfer vehicles, removing space debris, and providing power (from earth) to a lunar base during the long lunar night.

Lipinski, R.J.; McArthur, D.A. [Sandia National Labs., Albuquerque, NM (United States). Nuclear Systems Research

1994-10-01T23:59:59.000Z

204

A MOLTEN SALT NATURAL CONVECTION REACTOR SYSTEM  

SciTech Connect

Fuel-salt volumes external to the core of a molten-salt reactor are calculated for a system in which the fuel salt circulates through the core and primary exchanger by free convection. In the calculation of these volumes, the exchanger heights above the core top range from 5 to 20 ft. Coolants considered for the primary exchanger are a second molten salt and helium. External fuel holdup is found to be the same with either coolant. Two sets of terminal temperatures are selected for the helium. The first combination permits steam generation at 850 psia, 900 deg F. The second set is selected for a closed gas turbine cycle with an 1100 deg F turbine inlet temperature. Specific power (thermal kw/kg 235) is found to be about 900 Mv/kg, based on initial, clean conditions and a 60 Mw (thermal) output. A specific power of 1275 kw/kg is estimated for a forced convection system of the same rating. (auth)

Romie, F.E.; Kinyon, B.W.

1958-02-01T23:59:59.000Z

205

Reactor safety method  

DOE Patents (OSTI)

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

Vachon, Lawrence J. (Clairton, PA)

1980-03-11T23:59:59.000Z

206

NEUTRONIC REACTOR MANIPULATING DEVICE  

DOE Patents (OSTI)

A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

Ohlinger, L.A.

1962-08-01T23:59:59.000Z

207

Cold Vacuum Drying (CVD) Facility General Service Helium System Design Description  

DOE Green Energy (OSTI)

The purpose of this System Design Description (SDD) is to describe the characteristics of the Cold Vacuum Drying (CVD) Facility general service helium system. The general service helium system is a general service facility process support system, but does include safety-class structures, systems and components (SSCs) providing protection to the offsite public. The general service helium system also performs safety-significant functions that provide protection to onsite workers. The general helium system essential function is to provide helium (He) to support process functions during all phases of facility operations. General service helium is used to purge the cask and the MCO in order to maintain their internal atmospheres below hydrogen flammability concentrations. The general service helium system also supplies helium to purge the process water conditioning (PWC) lines and components and the vacuum purge system (VPS) vacuum pump. The general service helium system, if available following an Safety Class Instrument and Control System (SCIC) Isolation and Purge (IS0 and PURGE) Trip, can provide an alternate general service helium system source to supply the Safety-Class Helium (SCHe) System.

SHAPLEY, B.J.

2000-04-20T23:59:59.000Z

208

Commercial-Scale Performance Predictions for High-Temperature Electrolysis Plants Coupled to Three Advanced Reactor Types  

DOE Green Energy (OSTI)

This report presents results of system analyses that have been developed to assess the hydrogen production performance of commercial-scale high-temperature electrolysis (HTE) plants driven by three different advanced reactor – power-cycle combinations: a high-temperature helium cooled reactor coupled to a direct Brayton power cycle, a supercritical CO2-cooled reactor coupled to a direct recompression cycle, and a sodium-cooled fast reactor coupled to a Rankine cycle. The system analyses were performed using UniSim software. The work described in this report represents a refinement of previous analyses in that the process flow diagrams include realistic representations of the three advanced reactors directly coupled to the power cycles and integrated with the high-temperature electrolysis process loops. In addition, this report includes parametric studies in which the performance of each HTE concept is determined over a wide range of operating conditions. Results of the study indicate that overall thermal-to- hydrogen production efficiencies (based on the low heating value of the produced hydrogen) in the 45 - 50% range can be achieved at reasonable production rates with the high-temperature helium cooled reactor concept, 42 - 44% with the supercritical CO2-cooled reactor and about 33 - 34% with the sodium-cooled reactor.

M. G. McKellar; J. E. O'Brien; J. S. Herring

2007-09-01T23:59:59.000Z

209

THE INTEGRATION OF PROCESS HEAT APPLICATIONS TO HIGH TEMPERATURE GAS REACTORS  

SciTech Connect

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.

Michael G. McKellar

2011-11-01T23:59:59.000Z

210

An Engineering Test Reactor  

SciTech Connect

A relatively inexpensive reactor for the specific purpose of testing a sub-critical portion of another reactor under conditions that would exist during actual operation is discussed. It is concluded that an engineering tool for reactor development work that bridges the present gap between exponential and criticality experiments and the actual full scale operating reactor is feasible. An example of such a test reactor which would not entail development effort to ut into operation is depicted.

Fahrner, T.; Stoker, R.L.; Thomson, A.S.

1951-03-16T23:59:59.000Z

211

Assessment of HTGR Helium Compressor Analysis Tool Based on Newton-Raphson Numerical Application to Through-flow Analysis  

SciTech Connect

This study describes the development of a computer program for analyzing the off-design performance of axial flow helium compressors, which is one of the major concerns for the power conversion system of a high temperature gas-cooled reactor (HTGR). The compressor performance has been predicted by the aerodynamic analysis of meridional flow with allowances for losses. The governing equations have been derived from Euler turbomachine equation and the streamline curvature method, and then they have been merged into linearized equations based on the Newton-Raphson numerical method. The effect of viscosity is considered by empirical correlations to introduce entropy rises caused by primary loss sources. Use of the method has been illustrated by applying it to a 20-stage helium compressor of the GTHTR300 plant. As a result, the flow throughout the stages of the compressor has been predicted and the compressor characteristics have been also investigated according to the design specification. The program results show much better stability and good convergence with respect to other through-flow methods, and good agreement with the compressor performance map provided by JAEA. (authors)

Ji Hwan Kim; Hyeun Min Kim; Hee Cheon NO [Korea Advanced Institute of Science and Technology, 335 Gwahangno - 373-1 Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

2006-07-01T23:59:59.000Z

212

Reactor Pressure Vessel Task of Light Water Reactor Sustainability...  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure...

213

Memorandum on Chemical Reactors and Reactor Hazards  

SciTech Connect

Two important problems in the investigation of reactor hazards are the chemical reactivity of various materials employed in reactor construction and the chracteristics of heat transfer under transient conditions, specifically heat transfer when driven by an exponentially increasing heat source (exp t/T). Although these problems are independent of each other, when studied in relation to reactor hazards they may occur in a closely coupled sequence. For example the onset of a dangerous chemical reactor may be due to structural failure of various reactor components under an exponentially rising heat source originating with a runaway nuclear reactor. For this reason, these two problems should eventually be studied together after an exploratory experimental survey has been made in which they are considered separately.

Mills, M.M.; Pearlman, H.; Ruebsamen, W.; Steele, G., Chrisney, J.

1951-07-05T23:59:59.000Z

214

Si deposition rates in a two-dimensional CVD (chemical vapor deposition) reactor and comparisons with model calculations  

SciTech Connect

Deposition rates are presented for silicon from silane in a helium carrier gas using a tubular CVD reactor with a two-dimensional flow geometry. Measured surface-temperature profiles, inlet gas velocities, total pressures, and silane/helium concentrations are reported, providing exact boundary conditions that can be used in a two-dimensional numerical CVD model. Comparisons are made between this data and two variations of a model by Coltrin, Kee, and Miller in which different empirical expressions for the silane and disilane reactive sticking coefficient are used.

Breiland, W.G.; Coltrin, M.E.

1989-10-01T23:59:59.000Z

215

Metastable Helium Molecules as Tracers in Superfluid {sup 4}He  

SciTech Connect

Metastable helium molecules generated in a discharge near a sharp tungsten tip immersed in superfluid {sup 4}He are imaged using a laser-induced-fluorescence technique. By pulsing the tip, a small cloud of He{sub 2}* molecules is produced. We can determine the normal-fluid velocity in a heat-induced counterflow by tracing the position of a single molecule cloud. As we run the tip in continuous field-emission mode, a normal-fluid jet from the tip is generated and molecules are entrained in the jet. A focused 910 nm pump laser pulse is used to drive a small group of molecules to the first excited vibrational level of the triplet ground state. Subsequent imaging of the tagged molecules with an expanded 925 nm probe laser pulse allows us to measure the flow velocity of the jet. The techniques we developed provide new tools in quantitatively studying the normal fluid flow in superfluid helium.

Guo, W.; Wright, J. D.; Cahn, S. B.; Nikkel, J. A.; McKinsey, D. N. [Physics Department, Yale University, New Haven, Connecticut 06515 (United States)

2009-06-12T23:59:59.000Z

216

Shock Wave Structure for Argon, Helium, and Nitrogen  

E-Print Network (OSTI)

We compare the thickness of shock wave fronts at different Mach numbers, modeled via Navier-Stokes (NS) and Quasi-gasdynamic (QGD) equations, with experimental results from the literature. Monoatomic argon and helium, and diatomic nitrogen, are considered. In this modeling a finite-difference scheme with second-order spatial accuracy is employed. For argon the density thickness calculated via QGD and NS models are in good agreement with each other, and with the experimental results. For helium QGD and NS results agree well with those from the bimodal model. For nitrogen, the QGD results are closer to the experimental data than NS results. The QGD-based algorithm converges to the steady state solution faster than the NS-based one.

T. G. Elizarova; I. A. Shirokov; S. Montero

2004-07-28T23:59:59.000Z

217

Pulsed extraction of ionization from helium buffer gas  

E-Print Network (OSTI)

The migration of intense ionization created in helium buffer gas under the influence of applied electric fields is considered. First the chemical evolution of the ionization created by fast heavy-ion beams is described. Straight forward estimates of the lifetimes for charge exchange indicate a clear suppression of charge exchange during ion migration in low pressure helium. Then self-consistent calculations of the migration of the ions in the electric field of a gas-filled cell at the National Superconducting Cyclotron Laboratory (NSCL) using a Particle-In-Cell computer code are presented. The results of the calculations are compared to measurements of the extracted ion current caused by beam pulses injected into the NSCL gas cell.

D. J. Morrissey; G. Bollen; M. Facina; S. Schwarz

2008-08-13T23:59:59.000Z

218

A High Reliability Gas-driven Helium Cryogenic Centrifugal Compressor  

E-Print Network (OSTI)

A helium cryogenic compressor was developed and tested in real conditions in 1996. The achieved objective was to compress 0.018 kg/s Helium at 4 K @ 1000 Pa (10 mbar) up to 3000 Pa (30 mbar). This project was an opportunity to develop and test an interesting new concept in view of future needs. The main features of this new specific technology are described. Particular attention is paid to the gas bearing supported rotor and to the pneumatic driver. Trade off between existing technologies and the present work are presented with special stress on the bearing system and the driver. The advantages are discussed, essentially focused on life time and high reliability without maintenance as well as non pollution characteristic. Practical operational modes are also described together with the experimental performances of the compressor. The article concludes with a brief outlook of future work.

Bonneton, M; Gistau-Baguer, Guy M; Turcat, F; Viennot, P

1998-01-01T23:59:59.000Z

219

Helium-ion-induced release of hydrogen from graphite  

DOE Green Energy (OSTI)

The ion-induced release of hydrogen from AXF-5Q graphite was studied for 350-eV helium ions. The hydrogen was implanted into the graphite with a low energy (approx.200 eV) and to a high fluence. This achieved a thin (approx.10-nm), saturated near-surface region. The release of hydrogen was measured as a function of helium fluence. A model that includes ion-induced detrapping, retrapping, and surface recombination was used to analyze the experimental data. A value of (1.65 +- 0.2) x 10/sup -16/ cm/sup 2/ was obtained from the detrapping cross section, and a value of (0.5 to 4) x 10/sup -14/ cm/sup 4//atoms was obtained for the recombination coefficient. 11 refs., 4 figs.

Langley, R.A.

1987-01-01T23:59:59.000Z

220

Stimulated Brillouin Scattering from Helium-Hydrogen Plasmas  

DOE Green Energy (OSTI)

An extensive study of the stimulated Brillouin scattering (SBS) in helium-hydrogen plasmas has been performed using a gas jet at the Janus Laser Facility. We observe three regions of reflectivity by varying the probe intensity from 10{sup 14} to 10{sup 16}: saturated region, linear region, and near SBS threshold region. In the linear regime, adding small amounts of H to a He plasma reduces the SBS reflectivity by a factor of 4.

Froula, D H; Divol, L; Price, D; Gregori, G; Williams, E A; Glenzer, S H

2003-08-22T23:59:59.000Z

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


221

Semiclassical description of antiproton capture on atomic helium  

SciTech Connect

A semiclassical, many-body atomic model incorporating a momentum-dependent Heisenberg core to stabilize atomic electrons is used to study antiproton capture on helium. Details of the antiproton collisions leading to eventual capture are presented, including the energy and angular-momentum states of incident antiprotons which result in capture via single- or double-electron ionization, i.e., into He[sup 2+][ital [bar p

Beck, W.A. (Department of Physics, University of Washington, Seattle, Washington 98195 (United States) Quantum Medical Systems, Issaquah, Washington 98027 (United States)); Wilets, L. (Department of Physics, University of Washington, Seattle, Washington 98195 (United States)); Alberg, M.A. (Department of Physics, University of Washington, Seattle, Washington 98195 (United States) Department of Physics, Seattle University, Seattle, Washington 98122 (United States))

1993-10-01T23:59:59.000Z

222

Open Questions in Stellar Helium Burning Addressed With Real Photons  

E-Print Network (OSTI)

The outcome of helium burning is the formation of the two elements, carbon and oxygen. The ratio of carbon to oxygen at the end of helium burning is crucial for understanding the final fate of a progenitor star and the nucleosynthesis of heavy elements in Type II supernova, with oxygen rich star predicted to collapse to a black hole, and a carbon rich star to a neutron star. Type Ia supernovae (SNeIa) are used as standard candles for measuring cosmological distances with the use of an empirical light curve-luminosity stretching factor. It is essential to understand helium burning that yields the carbon/oxygen white dwarf and thus the initial stage of SNeIa. Since the triple alpha-particle capture reaction, $^{8}Be(\\alpha,\\gamma)^{12}C$, the first burning stage in helium burning, is well understood, one must extract the cross section of the $^{12}C(\\alpha,\\gamma)^{16}O$ reaction at the Gamow window (300 keV) with high accuracy of approximately 10% or better. This goal has not been achieved despite repeated strong statements that appeared in the literature. In particular constraint from the beta-delayed alpha-particle emission of $^{16}N$ were shown to not sufficiently restrict the p-wave cross section factor; e.g. a low value of $S_{E1}(300)$ can not be ruled out. Measurements at low energies, are thus mandatory for determining the elusive cross section factor for the $^{12}C(\\alpha,\\gamma)^{16}O$ reaction. We are constructing a Time Projection Chamber (TPC) for use with high intensity photon beams extracted from the HI$\\gamma$S/TUNL facility at Duke University to study the $^{16}O(\\gamma,\\alpha)^{12}C$ reaction, and thus the direct reaction at energies as low as 0.7 MeV. This work is in progress.

Moshe Gai

2003-03-20T23:59:59.000Z

223

Mathematical modeling of a Fermilab helium liquefier coldbox  

SciTech Connect

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.

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

1995-12-01T23:59:59.000Z

224

On the Evolution of Helium in Blue Compact Galaxies  

E-Print Network (OSTI)

We discuss the chemical evolution of dwarf irregular and blue compact galaxies in light of recent data, new stellar yields and chemical evolution models. We examine the abundance data for evidence of HII region self-enrichment effects, which would lead to correlations in the scatter of helium, nitrogen, and oxygen abundances around their mean trends. The observed helium abundance trends show no such correlations, though the nitrogen--oxygen trend does show strong evidence for real scatter beyond observational error. We construct simple models for the chemical evolution of these galaxies, using the most recent yields of \\he4, C, N and O in intermediate- and high-mass stars. The effects of galactic outflows, which can arise both from bulk heating and evaporation of the ISM, and from the partial escape of enriched supernova ejecta are included. In agreement with other studies, we find that supernova-enriched outflows can roughly reproduce the observed He, C, N, and O trends; however, in models that fit N versus O, the slopes $\\Delta Y/\\Delta$O and $\\Delta Y/\\Delta$N consistently fall more than $2\\sigma$ below the fit to observations. We discuss the role of the models and their uncertainties in the extrapolation of primordial helium from the data. We also explore the model dependence arising nucleosynthesis uncertainties associated with nitrogen yields in intermediate mass stars, the fate of $8-11 \\msol$ stars, and massive star winds.

Brian D. Fields; Keith A. Olive

1998-03-25T23:59:59.000Z

225

UCN production by multiphonon processes in superfluid Helium under pressure  

E-Print Network (OSTI)

Cold neutrons are converted to ultra-cold neutrons (UCN) by the excitation of a single phonon or multiphonons in superfluid helium. The dynamic scattering function S(q, omega) of the superfluid helium strongly depends on pressure, leading to a pressure- dependent differential UCN production rate. A phenomenological expression for the multiphonon part of the scattering function s(lambda) describing UCN production has been derived from inelastic neutron scattering data. When combined with the production rate from single phonon processes this allows us to calculate the UCN production for any incident neutron flux. For calculations of the UCN production from single phonon processes we propose to use the values for S*(SVP) = 0.118(8) and S*(20 bar) = 0.066(6). As an example we will calculate the expected UCN production rate at the cold neutron beam for fundamental physics PF1b at the Institut Laue Langevin. We conclude that UCN production in superfluid helium under pressure is not attractive.

P. Schmidt-Wellenburg; K. H. Andersen; O. Zimmer

2008-11-26T23:59:59.000Z

226

Prerequisite requirements for higher graphite temperature limits and/or nitrogen atmosphere at all reactors  

SciTech Connect

The graphite temperature limits specified in the process standards-are being closely approached or have limited power levels at B, D, DR, and reactors. An increase of approximately 50--100 C in graphite temperature limits at these reactors would permit year-around operation on bulk outlet temperature limits. There has been considerable recent interest in extending to all reactors the use of nitrogen as a reactor gas constituent. With present graphite temperature limits, significant benefit from nitrogen usage will not be obtained during the winter months at reactors which are graphite temperature limited with approximately 100 per cent helium. However, during the summer months when bulk outlet temperature limitations result in graphite temperatures considerably below the maximum permissible graphite temperatures, the substitution of nitrogen for carbon dioxide could be of value. Under these circumstances, both a reduction in helium usage and a reduction in enrichment costs could result. With an increase in permissible graphite temperature limits, year-around benefit from reduced helium usage and reduced enrichment costs would be possible. To meet the Pu-240 specification with higher graphite temperatures however, would require a reduction in current goal discharge exposures with resulting increased fuel and burnout costs. Additionally, the incentives for higher graphite temperatures are very sensitive to both the mode of operation and to the assumed product worth. An economic study by the Process Analysis Unit discussing in detail the factors influencing the incentive for higher graphite temperature will be published in the near future. At a meeting among the staff several weeks ago, the prerequisite requirements for permitting higher graphite temperatures and/or nitrogen usage at all reactors were discussed. It is the purpose of this letter to briefly outline the prerequisite steps considered necessary to achieve these goals.

Graves, S.M.

1962-02-26T23:59:59.000Z

227

Evaluation of High-Temperature Alloys for Helium Gas Turbines  

Science Conference Proceedings (OSTI)

C. 1. Mechanical Property / Status of Metallic Materials Development for Application in Advanced High-Temperature Gas-Cooled Reactor / Material

Wolfgang Jakobeit; Jörn-Peter Pfeifer; Georg Ullrich

228

Basic and Applied Science Research Reactors - Reactors designed...  

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

BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th...

229

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

SciTech Connect

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.

None

2010-09-01T23:59:59.000Z

230

J. Plasma Fusion Res. SERIES, Vol. 8 (2009) Alternative Fusion Reactors as Future Commercial Power Plants  

E-Print Network (OSTI)

Alternative reactor based on a field-reversed configuration (FRC) has advantages of the cylindrical geometry, the open field line geometry (direct energy conversion (DEC) of the charged-particle flow), and high ? (plasma pressure/magnetic-field pressure). This paper aims to evaluate the attractiveness of a low radioactive FRC fusion core. Analysis of a conceptual deuterium- helium-3 (D- 3 He) fusion power reactor is presented and reference point is defined. Principal parameters of the D- 3 He plasma reference case (RC) and comparison with conceptual D- 3 He tokamak and FRC power plants are shown.

Sergei V. Ryzhkov

2008-01-01T23:59:59.000Z

231

Attrition reactor system  

DOE Patents (OSTI)

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.

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

1993-01-01T23:59:59.000Z

232

Reactor Sharing Program  

Science Conference Proceedings (OSTI)

Progress achieved at the University of Florida Training Reactor (UFTR) facility through the US Department of Energy's University Reactor Sharing Program is reported for the period of 1991--1992.

Vernetson, W.G.

1993-01-01T23:59:59.000Z

233

Guidebook to nuclear reactors  

SciTech Connect

A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen.

Nero, A.V. Jr.

1976-05-01T23:59:59.000Z

234

Attrition reactor system  

DOE Patents (OSTI)

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.

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

1993-09-28T23:59:59.000Z

235

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

DOE Patents (OSTI)

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.

Golden, Timothy Christopher (Allentown, PA); Farris, Thomas Stephen (Bethlehem, PA)

2008-11-18T23:59:59.000Z

236

Helium Recovery in the LHC Cryogenic System following Magnet Resistive Transitions  

E-Print Network (OSTI)

A resistive transition (quench) of the Large Hadron Collider magnets provokes the expulsion of helium from the magnet cryostats to the helium recovery system. A high-volume, vacuum-insulated recovery line connected to several uninsulated medium-pressure gas storage tanks, forms the main constituents of the system. Besides a dedicated hardware configuration, helium recovery also implies specific procedures that should follow a quench, in order to conserve the discharged helium and possibly make use of its refrigeration capability. The amount of energy transferred after a quench from the magnets to the helium leaving the cold mass has been estimated on the basis of experimental data. Based on these data, the helium thermodynamic state in the recovery system is calculated using a lumped parameter approach. The LHC magnet quenches are classified ina parametric way from their cryogenic consequences and procedures that should follow the quench are proposed.

Chorowski, M; Serio, L; Tavian, L; Wagner, U; Van Weelderen, R

1998-01-01T23:59:59.000Z

237

NEUTRONIC REACTOR POWER PLANT  

DOE Patents (OSTI)

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)

Metcalf, H.E.

1962-12-25T23:59:59.000Z

238

High solids fermentation reactor  

DOE Patents (OSTI)

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.

Wyman, Charles E. (Lakewood, CO); Grohmann, Karel (Littleton, CO); Himmel, Michael E. (Littleton, CO); Richard, Christopher J. (Lakewood, CO)

1993-01-01T23:59:59.000Z

239

Improved vortex reactor system  

DOE Patents (OSTI)

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.

Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

1995-01-01T23:59:59.000Z

240

FAST NEUTRON REACTOR  

DOE Patents (OSTI)

A reactor comprising fissionable material in concentration sufficiently high so that the average neutron enengy within the reactor is at least 25,000 ev is described. A natural uranium blanket surrounds the reactor, and a moderating reflector surrounds the blanket. The blanket is thick enough to substantially eliminate flow of neutrons from the reflector.

Soodak, H.; Wigner, E.P.

1961-07-25T23:59:59.000Z

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


241

NUCLEAR REACTOR CONTROL SYSTEM  

DOE Patents (OSTI)

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.

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

1959-11-01T23:59:59.000Z

242

Very High Efficiency Reactor (VHER) Concepts for Electrical Power Generation and Hydrogen Production  

DOE Green Energy (OSTI)

The goal of the Very High Efficiency Reactor study was to develop and analyze concepts for the next generation of nuclear power reactors. The next generation power reactor should be cost effective compared to current power generation plant, passively safe, and proliferation-resistant. High-temperature reactor systems allow higher electrical generating efficiencies and high-temperature process heat applications, such as thermo-chemical hydrogen production. The study focused on three concepts; one using molten salt coolant with a prismatic fuel-element geometry, the other two using high-pressure helium coolant with a prismatic fuel-element geometry and a fuel-pebble element design. Peak operating temperatures, passive-safety, decay heat removal, criticality, burnup, reactivity coefficients, and material issues were analyzed to determine the technical feasibility of each concept.

PARMA JR.,EDWARD J.; PICKARD,PAUL S.; SUO-ANTTILA,AHTI JORMA

2003-06-01T23:59:59.000Z

243

Cold Vacuum Drying (CVD) Facility General Service Helium System Design Description  

DOE Green Energy (OSTI)

This document describes the Cold Vacuum Drying Facility general service helium system (GSHe). The GSHe is a general service facility process support system, but does include safety-class systems, structures and components providing protection to the offsite public. The GSHe also performs safety-significant functions that provide protection to onsite workers. The GSHe essential function is to provide helium to support process functions during all phases of facility operations. GSHe helium is used to purge the cask and the MCO in order to maintain their internal atmospheres below hydrogen flammability concentrations. The GSHe also supplies helium to purge the PWC lines and components and the VPS vacuum pump.

FARWICK, C.C.

1999-07-06T23:59:59.000Z

244

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

SciTech Connect

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.

Ely, James H.

2012-03-19T23:59:59.000Z

245

Ultra high vacuum pumping system and high sensitivity helium leak detector  

DOE Patents (OSTI)

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.

Myneni, G.R.

1997-12-30T23:59:59.000Z

246

Ultra high vacuum pumping system and high sensitivity helium leak detector  

DOE Patents (OSTI)

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.

Myneni, Ganapati Rao (Yorktown, VA)

1997-01-01T23:59:59.000Z

247

THE CONSTRUCTION AND TESTING OF A DOUBLE ACTING BELLOWS LIQUID HELIUM PUMP  

E-Print Network (OSTI)

Helium Pump (Without Heat Exchanger S Control Dewar) (LIQUIDcavities from the exit heat exchanger. Both of these valvescopper tube split flow heat exchanger which shares a common

Burns, W.A.

2010-01-01T23:59:59.000Z

248

The Argonaut Reactor - Reactors designed/built by Argonne National  

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

Achievements > Achievements > Argonne Reactors > Training Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

249

Nuclear reactor overflow line  

DOE Patents (OSTI)

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.

Severson, Wayne J. (Pittsburgh, PA)

1976-01-01T23:59:59.000Z

250

Reactor vessel support system  

DOE Patents (OSTI)

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.

Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

1982-01-01T23:59:59.000Z

251

Spinning fluids reactor  

SciTech Connect

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.

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

2012-11-20T23:59:59.000Z

252

Determining Reactor Neutrino Flux  

E-Print Network (OSTI)

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

Cao, Jun

2011-01-01T23:59:59.000Z

253

Reactor water cleanup system  

DOE Patents (OSTI)

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.

Gluntz, Douglas M. (San Jose, CA); Taft, William E. (Los Gatos, CA)

1994-01-01T23:59:59.000Z

254

Determining Reactor Neutrino Flux  

E-Print Network (OSTI)

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

Jun Cao

2011-01-12T23:59:59.000Z

255

Fission reactors and materials  

SciTech Connect

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

Frost, B.R.T.

1981-12-01T23:59:59.000Z

256

Helium measurements of pore-fluids obtained from SAFOD drillcore  

SciTech Connect

{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).

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

2010-04-15T23:59:59.000Z

257

A pencil beam algorithm for helium ion beam therapy  

Science Conference Proceedings (OSTI)

Purpose: To develop a flexible pencil beam algorithm for helium ion beam therapy. Dose distributions were calculated using the newly developed pencil beam algorithm and validated using Monte Carlo (MC) methods. Methods: The algorithm was based on the established theory of fluence weighted elemental pencil beam (PB) kernels. Using a new real-time splitting approach, a minimization routine selects the optimal shape for each sub-beam. Dose depositions along the beam path were determined using a look-up table (LUT). Data for LUT generation were derived from MC simulations in water using GATE 6.1. For materials other than water, dose depositions were calculated by the algorithm using water-equivalent depth scaling. Lateral beam spreading caused by multiple scattering has been accounted for by implementing a non-local scattering formula developed by Gottschalk. A new nuclear correction was modelled using a Voigt function and implemented by a LUT approach. Validation simulations have been performed using a phantom filled with homogeneous materials or heterogeneous slabs of up to 3 cm. The beams were incident perpendicular to the phantoms surface with initial particle energies ranging from 50 to 250 MeV/A with a total number of 10{sup 7} ions per beam. For comparison a special evaluation software was developed calculating the gamma indices for dose distributions. Results: In homogeneous phantoms, maximum range deviations between PB and MC of less than 1.1% and differences in the width of the distal energy falloff of the Bragg-Peak from 80% to 20% of less than 0.1 mm were found. Heterogeneous phantoms using layered slabs satisfied a {gamma}-index criterion of 2%/2mm of the local value except for some single voxels. For more complex phantoms using laterally arranged bone-air slabs, the {gamma}-index criterion was exceeded in some areas giving a maximum {gamma}-index of 1.75 and 4.9% of the voxels showed {gamma}-index values larger than one. The calculation precision of the presented algorithm was considered to be sufficient for clinical practice. Although only data for helium beams was presented, the performance of the pencil beam algorithm for proton beams was comparable. Conclusions: The pencil beam algorithm developed for helium ions presents a suitable tool for dose calculations. Its calculation speed was evaluated to be similar to other published pencil beam algorithms. The flexible design allows easy customization of measured depth-dose distributions and use of varying beam profiles, thus making it a promising candidate for integration into future treatment planning systems. Current work in progress deals with RBE effects of helium ions to complete the model.

Fuchs, Hermann; Stroebele, Julia; Schreiner, Thomas; Hirtl, Albert; Georg, Dietmar [Christian Doppler Laboratory for Medical Radiation Research for Radiation Oncology, Medical University of Vienna, 1090 Vienna (Austria); Department of Radiation Oncology, Medical University of Vienna/AKH Vienna, 1090 Vienna (Austria) and Comprehensive Cancer Center, Medical University of Vienna/AKH Vienna, 1090 Vienna (Austria); Department of Radiation Oncology, Medical University of Vienna/AKH Vienna (Austria) and Comprehensive Cancer Center, Medical University of Vienna/AKH Vienna, 1090 Vienna (Austria); PEG MedAustron, 2700 Wiener Neustadt (Austria); Department of Nuclear Medicine, Medical University of Vienna, 1090 Vienna (Austria); Christian Doppler Laboratory for Medical Radiation Research for Radiation Oncology, Medical University of Vienna, 1090 Vienna (Austria); Department of Radiation Oncology, Medical University of Vienna/AKH Vienna, 1090 Vienna (Austria) and Comprehensive Cancer Center, Medical University of Vienna/AKH Vienna, 1090 Vienna (Austria)

2012-11-15T23:59:59.000Z

258

Gas Diffusion in Metals: Fundamental Study of Helium-Point Defect Interactions in Iron and Kinetics of Hydrogen Desorption from Zirconium Hydride  

E-Print Network (OSTI)

Gas Diffusion in Metals: Fundamental Study of Helium-PointGas Diffusion in Metals: Fundamental Study of Helium-Point138 8.1.1 Fundamental study of helium-point defect

Hu, Xunxiang

2013-01-01T23:59:59.000Z

259

HEAVY-WATER-MODERATED POWER REACTORS ENGINEERING AND ECONOMIC EVALUATIONS. VOLUME I. SUMMARY REPORT  

SciTech Connect

Capital investments and the cost of power were estimated for 21 heavy- water-moderated, natural-uraniumfueled power-reactor plants, ranging in capacity from 100 to 460 Mw(e). Comparisons were made of hot- and coldmoderator reactors and of the relative merits of pressuretube and pressure-vessel designs. Reactors cooled with liquid D/sub 2/O, boiling D/sub 2/O, /sub 2/O steam, and helium were evalunted. A cold-moderator pressure-tube reactor cooled with boiling D/sub 2/O shows the most economic promise of the D/sub 2/Omoderated reactor systems studied to date. Reactors of this type have sufficient reactivity to permit satisfactory fuel exposures, but the development of additional technology is a prerequisite for optimum designs. At capacities of 300 and 400 Mw(e), the estimated power costs from the current designs of boiling-D/sub 2/O pressure-tabe reactor plants are 11.3 and 9.8 mills/kwh, respectively. From liquid-D/sub 2/-cooled concepts of comparable capacities the indicated power costs are 7 to 20% higher. With an active development program, a power cost of 8.0 to 8.5 mills/kwh may be attained in a 300 Mw(e) boiling-D/sub 2/O reactor plant within the next decade. (auth)

1960-06-01T23:59:59.000Z

260

DOE reactor-pumped laser program  

SciTech Connect

FALCON is a high-power, steady-state, nuclear reactor-pumped laser (RPL) concept that is being developed by the Department of Energy. The FALCON program has experimentally demonstrated reactor-pumped lasing in various mixtures of xenon, argon, neon, and helium at at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. The major strengths of a reactor-pumped laser are continuous high-power operation, modular construction, self-contained power, compact size, and a variety of wavelengths (from visible to infrared). These characteristics suggest numerous applications not easily accessible to other laser types. A ground-based RPL could beam its power to space for such activities as illuminating geosynchronous communication satellites in the earth`s shadow to extend their lives, beaming power to orbital transfer vehicles, removing space debris, and providing power (from earth) to a lunar base during the long lunar night. The compact size and self-contained power also makes an RPL very suitable for ship basing so that power-beaming activities could be situated around the globe. The continuous high power of an RPL opens many potential manufacturing applications such as deep-penetration welding and cutting of thick structures, wide-area hardening of metal surfaces by heat treatment or cladding application, wide-area vapor deposition of ceramics onto metal surfaces, production of sub-micron sized particles for manufacturing of ceramics, wide-area deposition of diamond-like coatings, and 3-D ceramic lithography.

Felty, J.R. [USDOE, Germantown, MD (United States). Defense Programs; Lipinski, R.J.; McArthur, D.A.; Pickard, P.S. [Sandia National Labs., Albuquerque, NM (United States)

1993-12-31T23:59:59.000Z

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


261

The Solar Wind Helium Abundance: Variation with Wind Speed and the Solar Cycle  

E-Print Network (OSTI)

The Solar Wind Helium Abundance: Variation with Wind Speed and the Solar Cycle Matthias R. Aellig Alamos National Lab., Los Alamos, NM 87545 Abstract We investigate the helium abundance in the solar wind of 1994 and early 2000 are analyzed. In agreement with similar work for previous solar cycles, we find

Richardson, John

262

THERMAL PERFORMANCE OF A DUAL-CHANNEL, HELIUM-COOLED, TUNGSTEN HEAT EXCHANGER  

E-Print Network (OSTI)

THERMAL PERFORMANCE OF A DUAL-CHANNEL, HELIUM-COOLED, TUNGSTEN HEAT EXCHANGER Dennis L. Youchison-cooled, refractory heat exchangers are now under consideration for first wall and divertor applications-channel, helium-cooled heat exchanger made almost entirely of tungsten was designed and fabricated by Thermacore

California at Los Angeles, University of

263

CONCEPTUAL DESIGN OF THE PEBBLE BED REACTOR EXPERIMENT  

SciTech Connect

The Pebble-Bed Reactor Experiment (PBRE) was designed to advance the pebble-bed concept by providing a test of characteristic features and make contriliutions to the general development of all-ceramic gas-cooled reactors. The following objectives were established for the reactor experiment: to investigate key features of the pebble-bed concept, including on-stream fuel handling, movement of fuel through bed, and performance of core; to obtain operation and maintenance experience with a system contaminated with fission- product activity; and to investigate the behavior of graphite fuel elements. A fourth objective, study of the behavior of core materials at conditions occurring with exit gas temperatures in the range 2000 to 2500 deg F, was tentatively included. The preliminary design oE a 5-Mw(t) reactor for achieving these objectives was prepared. The core of the PBRE is a 2 1/2-ft-diam, 4-ft-tall cylinder containing approximately 12,000 spherical graphite fuel elements 1 1/2 in. in diameter. Fuel spheres are added to and removed from the core by gravity flow, and these operations are performed while the reactor is at power by using pairs of valves for passage of elements into and out of the high-pressure system. Exposed fuel can be recycled to the top of the core. Helium coolant at 500 psia enters the bottom of the core at 550 deg F and emerges from the top at 1250 deg F. Concentric ducting connects the reactor to a single heat exchanger, which is located sufficiently high above the core that natural circulation will suffice to remove afterheat in the event the blower ceases to function. The coolant flow path is such that the entire pressure envelope is swept with helium at the temperature at which it emerges from the heat exchanger. Provision for semi- remote maintenance of contaminated components is emphasized in the layout, and most of the equipment in the primary and auxiliary systems is accessible from above by the removal of modular shielding units. Thc design permits replacement of the entire core graphite structure, The reactor can be adapted for testing core materials at high temperature by attemperation of the hot helium emerging from the core wwiih cool gas in a plenum in the upper graphite structure. Location of the PBRE at the site of the HRE-2 facility is proposed to take advantage of available buildings and services, but the reactor and auxiliary equipment will be contained in a completely new vessel located adjacent to the existing building. The design and direct construction cost of the reactor plant is estimated to be 958,000, allowance for contingencies, overhead, and escalation brings the total to ,260,000. High-temperature operation can be achieved when desired for an additional expenditure of less than 0,000. (auth)

1962-05-17T23:59:59.000Z

264

Gas Diffusion in Metals: Fundamental Study of Helium-Point Defect Interactions in Iron and Kinetics of Hydrogen Desorption from Zirconium Hydride  

E-Print Network (OSTI)

Ullmaier. Helium and hydrogen effects on the embrittlementT. Troev. The effect of hydrogen and helium on microvoidp. 1-51. M. Steinbruck. Hydrogen absorption by zirconium

Hu, Xunxiang

2013-01-01T23:59:59.000Z

265

Neutronic and thermal design considerations for heat-pipe reactors  

SciTech Connect

SABRE (Space-Arena Baseline Reactor) is a 100-kW/sub e/, heat-pipe-cooled, beryllium-reflected, fast reactor that produces heat at a temperature of 1500/sup 0/K and radiatively transmits it to high-temperature thermoelectric (TE) conversion elements. The use of heat pipes for core heat removal eliminates single-point failure mechanisms in the reactor cooling system, and provides minimal temperature drop radiative coupling to the TE array, as well as automatic, self-actuating removal of reactor afterheat. The question of how the failure of a fuel module heat pipe will affect neighboring fuel modules in the core is discussed, as is fission density peaking that occurs at the core/reflector interface. Results of neutronic calculations of the control margin available are described. Another issue that is addressed is that of helium generation in the heat pipes from neutron reactions in the core with the heat pipe fluid. Finally, the growth potential of the SABRE design to much higher powers is examined.

Ranken, W.A.; Koenig, D.R.

1983-01-01T23:59:59.000Z

266

Generation -IV Reactor Concepts  

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

Generation-IV Reactor Concepts Generation-IV Reactor Concepts Thomas H. Fanning Argonne National Laboratory 9700 South Cass Avenue Argonne, Illinois 60439, USA The Generation-IV International Forum (GIF) is a multi-national research and development (R&D) collaboration. The GIF pursues the development of advanced, next generation reactor technology with goals to improve: a) sustainability (effective fuel utilization and minimization of waste) b) economics (competitiveness with respect to other energy sources) c) safety and reliability (e.g., no need for offsite emergency response), and d) proliferation resistance and physical protection The GIF Technology Roadmap exercise selected six generic systems for further study: the Gas- cooled Fast Reactor (GFR), the Lead-cooled Fast Reactor (LFR), the Molten Salt Reactor (MSR),

267

HORIZONTAL BOILING REACTOR SYSTEM  

DOE Patents (OSTI)

Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

Treshow, M.

1958-11-18T23:59:59.000Z

268

Regional And Local Trends In Helium Isotopes, Basin And Range Province,  

Open Energy Info (EERE)

And Local Trends In Helium Isotopes, Basin And Range Province, And Local Trends In Helium Isotopes, Basin And Range Province, Western North America- Evidence For Deep Permeable Pathways Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Conference Paper: Regional And Local Trends In Helium Isotopes, Basin And Range Province, Western North America- Evidence For Deep Permeable Pathways Details Activities (1) Areas (1) Regions (0) Abstract: Fluids from the western margin of the Basin and Range have helium isotope ratios as high as ~6-7 Ra, indicating a strong mantle melt influence and consistent with recent and current volcanic activity. Moving away from these areas, helium isotope ratios decrease rapidly to 'background' values of around 0.6 Ra, and then gradually decrease toward the east to low values of ~0.1 Ra at the eastern margin of the Basin and

269

Unique method for liquid nitrogen precooling of a plate fin heat exchanger in a helium refrigeration cycle  

E-Print Network (OSTI)

Unique method for liquid nitrogen precooling of a plate fin heat exchanger in a helium refrigeration cycle

Weber, T B; Howell, G; Racine, M; Weisend, J G

2004-01-01T23:59:59.000Z

270

THERMAL NEUTRONIC REACTOR  

DOE Patents (OSTI)

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

Spinrad, B.I.

1960-01-12T23:59:59.000Z

271

Improved vortex reactor system  

DOE Patents (OSTI)

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.

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

1995-05-09T23:59:59.000Z

272

NUCLEAR REACTORS AND EARTHQUAKES  

SciTech Connect

A book is presented which supplies pertinent seismological information to engineers in the nuclear reactor field. Data are presented on the occurrence, intensity, and wave shapes. Techniques are described for evaluating the response of structures to such events. Certain reactor types and their modes of operation are described briefly. Various protection systems are considered. Earthquake experience in industrial and reactor plants is described. (D.L.C.)

1961-01-01T23:59:59.000Z

273

Pressurized fluidized bed reactor  

DOE Patents (OSTI)

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.

Isaksson, Juhani (Karhula, FI)

1996-01-01T23:59:59.000Z

274

Pressurized fluidized bed reactor  

DOE Patents (OSTI)

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.

Isaksson, J.

1996-03-19T23:59:59.000Z

275

Tokamak reactor first wall  

DOE Patents (OSTI)

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.

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

1984-11-20T23:59:59.000Z

276

Advanced Nuclear Research Reactor  

SciTech Connect

This report describes technical modifications implemented by INVAP to improve the safety of the Research Reactors the company designs and builds.

Lolich, J.V.

2004-10-06T23:59:59.000Z

277

HOMOGENEOUS NUCLEAR POWER REACTOR  

DOE Patents (OSTI)

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.

King, L.D.P.

1959-09-01T23:59:59.000Z

278

NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) PROGRAM GRANT NUMBER DE-FG03-00SF22168 TECHNICAL PROGRESS REPORT (Nov. 15, 2001 - Feb. 15,2002) ''Design and Layout Concepts for Compact, Factory-Produced, Transportable, Generation IV Reactor Systems''  

SciTech Connect

The objectives of this project are to develop and evaluate nuclear power plant designs and layout concepts to maximize the benefits of compact modular Generation IV reactor concepts including factory fabrication and packaging for optimal transportation and siting. Three nuclear power plant concepts are being studied representing water, helium and lead-bismuth coolants. This is the sixth quarterly progress report.

Fred R. Mynatt; Andy Kadak; Marc Berte; Larry Miller; Mohammed Khan; Joe McConn; Lawrence Townsend; Wesley Williams; Martin Williamson

2002-03-15T23:59:59.000Z

279

On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks  

SciTech Connect

This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

Samuel Bays; Ayodeji Alajo

2010-05-01T23:59:59.000Z

280

Foreign Research Reactor/Domestic Research Reactor Receipt Coordinator...  

National Nuclear Security Administration (NNSA)

Research ReactorDomestic Research Reactor Receipt Coordinator, Savannah River Nuclear Solutions | National Nuclear Security Administration Our Mission Managing the Stockpile...

Note: This page contains sample records for the topic "helium reactor gt-mhr" from the National Library of EnergyBeta (NLEBeta).
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281

ADMINISTRATION OF ORNL RESEARCH REACTORS  

SciTech Connect

Organization of the ORNL Operations division for administration of the Oak Ridge Research Reactor, the Low Intensity Testing Reactor, and the Oak Ridge Graphite Reactor is described. (J.R.D.)

Casto, W.R.

1962-08-20T23:59:59.000Z

282

Pion Electroproduction form Helium 3, Deuterium, and Hydrogen  

DOE Green Energy (OSTI)

A series of measurements for pion electroproduction from helium-3, deuterium, and hydrogen were completed at the Thomas Jefferson National Accelerator Facility by the NucPi Collaboration. E91003 began taking data in February 1998 and was completed in April 1998. The longitudinal and transverse parts of the differential cross section were extracted, by means of a Rosenbluth type separation, in the direction parallel to the virtual photon, at Q 2 = 0.4 GeV 2 , for W = 1.15 and W = 1.6 GeV. The mass dependence of the longitudinal cross section should provide insight into the surprising apparent absence of any significant cross section enhancement due to excess pions in the nuclear medium.

S. Avery

2002-05-01T23:59:59.000Z

283

Ionization of hydrogen and ionized helium by slow antiprotons  

E-Print Network (OSTI)

We study the ionization process involving antiproton and hydrogen in the energy range between 0.1 keV to 500 keV, using single center close coupling approximation. We construct the scattering wave function using B-spline bases. The results obtained for ionization of atomic hydrogen are compared with other existing theoretical calculations as well as with the available experimental data. The present results are found to be encouraging. We also employed this method to study the ionization of ionized helium in the energy range between 1 and 500 keV. On comparision, the present results are found to interpret well the cross section values calculated using other theories.

Sahoo, S; Walters, H R J

2004-01-01T23:59:59.000Z

284

Nuclear fusion in muonic deuterium-helium complex  

E-Print Network (OSTI)

Experimental study of the nuclear fusion reaction in charge-asymmetrical d-mu-3He complex is presented. The 14.6 MeV protons were detected by three pairs of Si(dE-E) telescopes placed around the cryogenic target filled with the deuterium + helium-3 gas at 34 K. The 6.85 keV gamma rays emitted during the de-excitation of d-mu-3He complex were detected by a germanium detector. The measurements were performed at two target densities, 0.0585 and 0.169 (relative to liquid hydrogen density) with an atomic concentration of 3He c=0.0469. The values of the effective rate of nuclear fusion in d-mu-3He was obtained for the first time, and the J=0 nuclear fusion rate in d-mu-3He was derived.

V. M. Bystritsky; M. Filipowicz; V. V. Gerasimov; P. E. Knowles; F. Mulhauser; N. P. Popov; V. A. Stolupin; V. P. Volnykh; J. Wozniak

2005-06-22T23:59:59.000Z

285

Production reactor characteristics  

SciTech Connect

Reactors for the production of special nuclear materials share many similarities with commercial nuclear power plants. Each relies on nuclear fission, uses uranium fuel, and produces large quantities of thermal power. However, there are some important differences in production reactor characteristics that may best be discussed in terms of mission, role, and technology.

Thiessen, C.W.; Hootman, H.E.

1990-01-01T23:59:59.000Z

286

Advanced converter reactors  

SciTech Connect

Advanced converter reactors (ACRs) of primary US interest are those which can be commercialized within about 20 years, and are: Advanced Light-Water Reactors, Spectral-Shift-Control Reactors, Heavy-Water Reactors (CANDU type), and High-Temperature Gas-Cooled Reactors. These reactors can operate on uranium, thorium, or uranium-thorium fuel cycles, but have the greatest fuel utilization on thorium type cycles. The water reactors tend to operate more economically on uranium cycles, while the HTGR is more economical on thorium cycles. Thus, the HTGR had the greatest practical potential for improving fuel utilization. If the US has 3.4 to 4 million tons U/sub 3/O/sub 8/ at reasonable costs, ACRs can make important contributions to maintaining a high nuclear power level for many decades; further, they work well with fast breeder reactors in the long term under symbiotic fueling conditions. Primary nuclear data needs of ACRs are integral measurements of reactivity coefficients and resonance absorption integrals.

Kasten, P.R.

1979-01-01T23:59:59.000Z

287

NEUTRONIC REACTOR SYSTEM  

DOE Patents (OSTI)

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.

Treshow, M.

1959-02-10T23:59:59.000Z

288

NEUTRONIC REACTOR BURIAL ASSEMBLY  

DOE Patents (OSTI)

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.

Treshow, M.

1961-05-01T23:59:59.000Z

289

The Integral Fast Reactor  

SciTech Connect

Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs.

Till, C.E.; Chang, Y.I. (Argonne National Lab., IL (USA)); Lineberry, M.J. (Argonne National Lab., Idaho Falls, ID (USA))

1990-01-01T23:59:59.000Z

290

Superconducting cable cooling system by helium gas at two pressures  

DOE Patents (OSTI)

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.

Dean, John W. (Los Alamos, NM)

1977-01-01T23:59:59.000Z

291

Helium measurements of pore-fluids obtained from SAFOD drillcore  

Science Conference Proceedings (OSTI)

{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 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).

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

2010-04-15T23:59:59.000Z

292

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

SciTech Connect

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.

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.

2003-11-12T23:59:59.000Z

293

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT  

SciTech Connect

Progress on reactor programs and in general engineering research and development programs is summarized. Research and development are reported on water-cooled reactors including EBWR and Borax-V, sodium-cooled reactors including ZPR-III, IV, and IX, Juggernaut, and EBR-I and II. Other work included a review of fast reactor technology, and studies on nuclear superheat, thermal and fast reactor safety, and reactor physics. Effort was also devoted to reactor materials and fuels development, heat engineering, separation processes and advanced reactor concepts. (J.R.D.)

1961-04-01T23:59:59.000Z

294

HEAVY-WATER-MODERATED POWER REACTORS ENGINEERING AND ECONOMIC EVALUATIONS. VOLUME II. ENGINEERING STUDIES AND TECHNICAL DATA  

SciTech Connect

The results of preliminary design and evaluation studies of various concepts of a power reactor that is moderated by heavy water and fueled with natural uranium are presented. Twenty-nine conceptal designs were developed for reactors ranging in capacity from 100 Mwe to 460 Mwe. Resigns were prepared for hot- and cold-moderator reac tors of the pressure vessel type, with liquid D/sub 2/O, boiling D/sub 2/O, E/sub 2/O steam, and helium as coolants. Also studied were cold-moderator pressure tube reactors cooled with liquid D/sub 2/O and boiling D/sub 2/O. The repont includes the results of engineering studies of the reactor systems, electrical generation facilities, and auxiliary equipment. (auth)

1960-12-01T23:59:59.000Z

295

Apparatus to measure liquid helium boil-off from low-loss superconducting current leads  

DOE Green Energy (OSTI)

A low-loss liquid helium dewar was constructed to measure the liquid helium boil-off rate from high-temperature superconducting current leads. The dewar has a measured background heat leakage rate of 12 mW. Equations calculating the heat leakage rate from the measured vapor mass flow rate in liquid helium boil-off experiments are derived. Parameters that affect the experiments, such as density ratio, absolute pressure, and rate of pressure variation, are discussed. This study is important as superconducting current leads may be used in superconducting magnetic energy storage systems.

Cha, Y.S.; Niemann, R.C.; Hull, J.R. [Argonne National Lab., IL (United States). Energy Technology Div.

1995-06-01T23:59:59.000Z

296

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

Science Conference Proceedings (OSTI)

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.

Maziasz, P.J.

1985-11-01T23:59:59.000Z

297

Forced flow supercritical helium in a closed heat transfer loop subjected to pulsed heat loads  

Science Conference Proceedings (OSTI)

The superconducting magnets of the tokamak JT-60SA are cooled by means of forced flows of supercritical helium at 4.4 K and 0.5 MPa. The closed loops transfer heat from the magnets to the refrigerator through heat exchangers immersed into a saturated liquid helium bath. An experimental loop was designed to represent a 1/20 scaled down mock-up of JT-60SA central solenoid cooling circuits. This design for keeping the same transit times in the helium circuits

2012-01-01T23:59:59.000Z

298

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

DOE Green Energy (OSTI)

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.

Shanahan, K; Jeffrey Holder, J

2006-07-10T23:59:59.000Z

299

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

DOE Green Energy (OSTI)

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.

Shanahan, K; Jeffrey Holder, J

2006-08-17T23:59:59.000Z

300

Nuclear reactor control column  

DOE Patents (OSTI)

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.

Bachovchin, Dennis M. (Plum Borough, PA)

1982-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "helium reactor gt-mhr" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
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to obtain the most current and comprehensive results.


301

Slurry reactor design studies  

SciTech Connect

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.

Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

1990-06-01T23:59:59.000Z

302

Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)  

DOE Green Energy (OSTI)

A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the equivalent temperature of heat delivered to either the power conversion system or a hydrogen production plant. Using a comparative cost analysis, the construction costs per unit output are projected to be 50-55% of the costs for modular gas-cooled or sodium-cooled reactor systems. This is primarily a consequence of substantially larger power output and higher conversion efficiency for the AHTR. The AHTR has a number of unique technical challenges in meeting the NGNP requirements; however, it appears to offer advantages over high-temperature helium-cooled reactors and provides an alternative development path to achieve the NGNP requirements. Primary challenges include optimizing the core design for improved response to transients, designing an internal blanket to thermally protect the reactor vessel, and engineering solutions to high-temperature refueling and maintenance.

Ingersoll, D.T.

2004-07-29T23:59:59.000Z

303

Nuclear reactor reflector  

DOE Patents (OSTI)

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.

Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

1994-01-01T23:59:59.000Z

304

Nuclear reactor reflector  

DOE Patents (OSTI)

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.

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

1994-06-07T23:59:59.000Z

305

Microfluidic electrochemical reactors  

DOE Patents (OSTI)

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.

Nuzzo, Ralph G. (Champaign, IL); Mitrovski, Svetlana M. (Urbana, IL)

2011-03-22T23:59:59.000Z

306

Fast Breeder Reactor studies  

Science Conference Proceedings (OSTI)

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.

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

1980-07-01T23:59:59.000Z

307

NUCLEAR REACTOR FUEL SYSTEMS  

DOE Patents (OSTI)

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.

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

1959-09-15T23:59:59.000Z

308

Spherical torus fusion reactor  

DOE Patents (OSTI)

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.

Martin Peng, Y.K.M.

1985-10-03T23:59:59.000Z

309

CONTROL FOR NEUTRONIC REACTOR  

DOE Patents (OSTI)

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.

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

1959-03-31T23:59:59.000Z

310

MHD power conversion system for NERVA reactor  

Science Conference Proceedings (OSTI)

Optimized linear magnetohydrodynamic (MHD) systems to produce 200 MWe for 1000 seconds are defined. For the specific mission envisioned, a mass flow of 45.36 Kg/sec (100 lbs/sec) of hydrogen is available to the system. Westinghouse NERVA reactor technology can heat this mass flow of hydrogen to 2550 K at a pressure of 12 atm. This hydrogen flow is assumed to be seeded with cesium to obtain the required MHD generator conductivity. For each MHD system concept considered, the MHD generator design was optimized in terms of operating Mach number, load parameter, and cesium seed fraction. The simplest concept, an open-cycle MHD system, is optimized by minimizing the total magnet plus cesium seed mass. The resulting magnet and magnet plus seed mass are 44092 and 54143 Kg respectively. The second concept considered was an open-cycle MHD system with seed recovery and reuse. It is optimized by minimizing the magnet mass. The resulting magnet mass is 40110 Kg. A third concept, a closed-cycle MHD system, was also considered. If only equilibrium conductivity is considered, cesium seeded hydrogen is shown to be the more attractive than cesium seeded helium, and the optimum generator would be identical to that for the open-cycle MHD system with seed recovery.

Seikel, G.R.; Condit, W.C.

1985-01-01T23:59:59.000Z

311

Helium isotopes in geothermal systems- Iceland, The Geysers, Raft River and  

Open Energy Info (EERE)

Helium isotopes in geothermal systems- Iceland, The Geysers, Raft River and Helium isotopes in geothermal systems- Iceland, The Geysers, Raft River and Steamboat Springs Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: Helium isotopes in geothermal systems- Iceland, The Geysers, Raft River and Steamboat Springs Details Activities (3) Areas (3) Regions (0) Abstract: Helium isotope ratios have been measured in geothermal fluids from Iceland, The Geysers, Raft River, Steamboat Springs and Hawaii. These ratios have been interpreted in terms of the processes which supply He in distinct isotopic ratios (i.e. magmatic He, ~10 Ra; atmospheric He, R,sub>a; and crustal He, ~0.1 Ra) and in terms of the processes which can alter the isotopic ratio (hydrologic mixing, U-Th series alpha production and weathering release of crustal He, magma aging and

312

Helium Isotopes In Geothermal And Volcanic Gases Of The Western United  

Open Energy Info (EERE)

Helium Isotopes In Geothermal And Volcanic Gases Of The Western United Helium Isotopes In Geothermal And Volcanic Gases Of The Western United States, I, Regional Variability And Magmatic Origin Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: Helium Isotopes In Geothermal And Volcanic Gases Of The Western United States, I, Regional Variability And Magmatic Origin Details Activities (1) Areas (1) Regions (0) Abstract: Helium isotope ratios in gases of thirty hot springs and geothermal wells and of five natural gas wells in the western United States show no relationship to regional conductive heat flow, but do show a correlation with magma-based thermal activity and reservoir fluid temperature (or total convective heat discharge). Gases from high-T (> 200°C) reservoirs have 3He/4He > 2 _ the atmospheric value, with high He

313

Flow of mantle fluids through the ductile lower crust: Helium isotope trends  

E-Print Network (OSTI)

super-imposed on regional trends can identify potentialCrust: Helium Isotope Trends I B. Mack Kennedyl andlMatthijsin the west, a regional trend that correlates with the rates

Kennedy, B. Mack; van Soest, Matthijs C.

2008-01-01T23:59:59.000Z

314

Helium and lead isotope geochemistry of oceanic volcanic rocks from the East Pacific and South Atlantic  

E-Print Network (OSTI)

The isotopic evolution of helium and lead in the Earth is coupled by virtue of their common radioactive parents uranium and thorium. The isotopic signatures in oceanic volcanic rocks provide constraints on the temporal ...

Graham, David W. (David William)

1987-01-01T23:59:59.000Z

315

Nanostructuring of molybdenum and tungsten surfaces by low-energy helium ions  

Science Conference Proceedings (OSTI)

The formation of metallic nanostructures by exposure of molybdenum and tungsten surfaces to high fluxes of low energy helium ions is studied as a function of the ion energy, plasma exposure time, and surface temperature. Helium plasma exposure leads to the formation of nanoscopic filaments on the surface of both metals. The size of the helium-induced nanostructure increases with increasing surface temperature while the thickness of the modified layer increases with time. In addition, the growth rate of the nanostructured layer also depends on the surface temperature. The size of the nanostructure appears linked with the size of the near-surface voids induced by the low energy ions. The results presented here thus demonstrate that surface processing by low-energy helium ions provides an efficient route for the formation of porous metallic nanostructures.

De Temmerman, Gregory; Bystrov, Kirill; Zielinski, Jakub J.; Balden, Martin; Matern, Gabriele; Arnas, Cecile; Marot, Laurent [FOM Institute DIFFER, Ducth Institute For Fundamental Energy Research, Association EURATOM-FOM, Trilateral Euregio Cluster, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Max Planck Institute for Plasma Physics, EURATOM Association, 85748 Garching (Germany); Laboratoire PIIM, CNRS/Aix-Marseille Universit, 13397 Marseille (France); Department of Physics, University of Basel, Basel, CH-4056 (Switzerland)

2012-07-15T23:59:59.000Z

316

Flow of mantle fluids through the ductile lower crust: Helium isotope trends  

E-Print Network (OSTI)

of the latter is the Dixie Valley thermal system (Figure 2),that crossed through Dixie Valley revealed a zone of lowsuggesting that the Dixie Valley helium anomaly rellects a

Kennedy, B. Mack; van Soest, Matthijs C.

2008-01-01T23:59:59.000Z

317

Helium/Hydrogen Effects on the Properties of Materials for the APT Target/Blanket Region  

DOE Green Energy (OSTI)

This paper provides a technical basis to rationalize potential synergistic effects among displacement damage and the hydrogen and helium embrittlement processes and suggests that such synergistic effects may be of significant importance to component performance in intense spallation neutron sources.

Louthan, M.R.

1999-11-17T23:59:59.000Z

318

Disk magnetohydrodynamic power conversion system for NERVA reactor  

Science Conference Proceedings (OSTI)

The combination of a magnetohydrodynamic (MHD) generator of the disk type with a NERVA reactor yields an advanced power system particularly suited to space applications with the capability of producing up to gigawatt pulses and multi-megawatt continuous operation. Several unique features result from the combination of this type of reactor and a disk MHD generator in which hydrogen serves as the plasma working fluid. Cesium seedings is utilized under conditions which enable the generator to operate stably in the non-equilibrium electrical conduction mode. In common with all practical MHD generators, the disk output is DC and voltages in the range 20--100 kV are attainable. This leads to a simplification of the power conditioning system and a major reduction in specific mass. Taken together with the high performance capabilities of the NERVA reactor, the result is an attractively low overall system specific mass. Further, the use of non-equilibrium ionization enables system specific enthalpy extractions in excess of 40% to be attained. This paper reports the results of a study to establish the basis for the design of a cesium seeded hydrogen MHD disk generator. Generator performance results are presented in terms of a stability factor which is related to cesium seeded hydrogen plasma behavior. It is shown that application of the results already obtained with cesium seeded noble gases (argon and helium) to the case of hydrogen as the working fluid in a disk MHD generator enables a high performance power system to be defined.

Jackson, W.D. (HMJ Corporation. 10400 Connecticut Ave., Kensington, Maryland 20895 (United States)); Bernard, F.E. (Westinghouse Corp., P.O. Box 355, Pittsburgh, Pennsylvania 15230 (United States)); Holman, R.R. (HMJ Corporation, 10400 Connecticut Ave., Kensington, Maryland 20895 (United States)); Maxwell, C.D. (STD Research Corp., P.O. Box C, Arcadia, California 91006 (United States)); Seikel, G.R. (SeiTec, Inc., P.O. Box 81264, Cleveland, Ohio 44181 (United States))

1993-01-15T23:59:59.000Z

319

Gas tagging and cover gas combination for nuclear reactor  

DOE Patents (OSTI)

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.

Gross, Kenny C. (Lemont, IL); Laug, Matthew T. (Idaho Falls, ID)

1985-01-01T23:59:59.000Z

320

Prognostics Health Management for Advanced Small Modular Reactor Passive Components  

SciTech Connect

In the United States, sustainable nuclear power to promote energy security is a key national energy priority. Advanced small modular reactors (AdvSMR), which are based on modularization of advanced reactor concepts using non-light-water reactor (LWR) coolants such as liquid metal, helium, or liquid salt may provide a longer-term alternative to more conventional LWR-based concepts. The economics of AdvSMRs will be impacted by the reduced economy-of-scale savings when compared to traditional LWRs and the controllable day-to-day costs of AdvSMRs are expected to be dominated by operations and maintenance costs. Therefore, achieving the full benefits of AdvSMR deployment requires a new paradigm for plant design and management. In this context, prognostic health management of passive components in AdvSMRs can play a key role in enabling the economic deployment of AdvSMRs. In this paper, the background of AdvSMRs is discussed from which requirements for PHM systems are derived. The particle filter technique is proposed as a prognostics framework for AdvSMR passive components and the suitability of the particle filter technique is illustrated by using it to forecast thermal creep degradation using a physics-of-failure model and based on a combination of types of measurements conceived for passive AdvSMR components.

Meyer, Ryan M.; Ramuhalli, Pradeep; Coble, Jamie B.; Mitchell, Mark R.; Wootan, David W.; Hirt, Evelyn H.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

2013-10-18T23:59:59.000Z

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


321

Carburization of austenitic alloys by gaseous impurities in helium  

SciTech Connect

The carburization behavior of Alloy 800H, Inconel Alloy 617 and Hastelloy Alloy X in helium containing various amounts of H/sub 2/, CO, CH/sub 4/, H/sub 2/O and CO/sub 2/ was studied. Corrosion tests were conducted in a temperature range from 649 to 1000/sup 0/C (1200 to 1832/sup 0/F) for exposure time up to 10,000 h. Four different helium environments, identified as A, B, C, and D, were investigated. Concentrations of gaseous impurities were 1500 ..mu..atm H/sub 2/, 450 ..mu..atm CO, 50 ..mu..atm CH/sub 4/ and 50 ..mu..atm H/sub 2/O for Environment A; 200 ..mu..atm H/sub 2/, 100 ..mu..atm CO, 20 ..mu..atm CH/sub 4/, 50 ..mu..atm H/sub 2/O and 5 ..mu..atm CO/sub 2/ for Environment B; 500 ..mu..atm H/sub 2/, 50 ..mu..atm CO, 50 ..mu..atm CH/sub 4/ and < 0.5 ..mu..atm H/sub 2/O for Environment C; and 500 ..mu..atm H/sub 2/, 50 ..mu..atm CO, 50 ..mu..atm CH/sub 4/ and 1.5 ..mu..atm H/sub 2/O for Environment D. Environments A and B were characteristic of high-oxygen potential, while C and D were characteristic of low-oxygen potential. The results showed that the carburization kinetics in low-oxygen potential environments (C and D) were significantly higher, approximately an order of magnitude higher at high temperatures, than those in high-oxygen potential environments (A and B) for all three alloys. Thermodynamic analyses indicated no significant differences in the thermodynamic carburization potential between low- and high-oxygen potential environments. It is thus believed that the enhanced carburization kinetics observed in the low-oxygen potential environments were related to kinetic effects. A qualitatively mechanistic model was proposed to explain the enhanced kinetics. The present results further suggest that controlling the oxygen potential of the service environment can be an effective means of reducing carburization of alloys.

Lai, G.Y.; Johnson, W.R.

1980-03-01T23:59:59.000Z

322

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

Science Conference Proceedings (OSTI)

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.

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

2008-04-01T23:59:59.000Z

323

Reactor hot spot analysis  

SciTech Connect

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.

Vilim, R.B.

1985-08-01T23:59:59.000Z

324

Reactor Thermal-Hydraulics  

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

Thermal-Hydraulics Thermal-Hydraulics Dr. Tanju Sofu, Argonne National Laboratory In a power reactor, the energy produced in fission reaction manifests itself as heat to be removed by a coolant and utilized in a thermodynamic energy conversion cycle to produce electricity. A simplified schematic of a typical nuclear power plant is shown in the diagram below. Primary coolant loop Steam Reactor Heat exchanger Primary pump Secondary pump Condenser Turbine Water Although this process is essentially the same as in any other steam plant configuration, the power density in a nuclear reactor core is typically four orders of magnitude higher than a fossil fueled plant and therefore it poses significant heat transfer challenges. Maximum power that can be obtained from a nuclear reactor is often limited by the

325

Compact power reactor  

DOE Patents (OSTI)

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.

Wetch, Joseph R. (Woodland Hills, CA); Dieckamp, Herman M. (Canoga Park, CA); Wilson, Lewis A. (Canoga Park, CA)

1978-01-01T23:59:59.000Z

326

Molten metal reactors  

SciTech Connect

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.

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

2013-11-05T23:59:59.000Z

327

THE GREEN BANK TELESCOPE H II REGION DISCOVERY SURVEY. IV. HELIUM AND CARBON RECOMBINATION LINES  

SciTech Connect

The Green Bank Telescope H II Region Discovery Survey (GBT HRDS) found hundreds of previously unknown Galactic regions of massive star formation by detecting hydrogen radio recombination line (RRL) emission from candidate H II region targets. Since the HRDS nebulae lie at large distances from the Sun, they are located in previously unprobed zones of the Galactic disk. Here, we derive the properties of helium and carbon RRL emission from HRDS nebulae. Our target sample is the subset of the HRDS that has visible helium or carbon RRLs. This criterion gives a total of 84 velocity components (14% of the HRDS) with helium emission and 52 (9%) with carbon emission. For our highest quality sources, the average {sup 4}He{sup +}/H{sup +} abundance ratio by number, (y {sup +}), is 0.068 {+-} 0.023(1{sigma}). This is the same ratio as that measured for the sample of previously known Galactic H II regions. Nebulae without detected helium emission give robust y {sup +} upper limits. There are 5 RRL emission components with y {sup +} less than 0.04 and another 12 with upper limits below this value. These H II regions must have either a very low {sup 4}He abundance or contain a significant amount of neutral helium. The HRDS has 20 nebulae with carbon RRL emission but no helium emission at its sensitivity level. There is no correlation between the carbon RRL parameters and the 8 {mu}m mid-infrared morphology of these nebulae.

Wenger, Trey V.; Bania, T. M. [Astronomy Department, 725 Commonwealth Avenue, Boston University, Boston, MA 02215 (United States)] [Astronomy Department, 725 Commonwealth Avenue, Boston University, Boston, MA 02215 (United States); Balser, Dana S. [National Radio Astronomy Observatory, 520 Edgemont Road, Charlottesville, VA, 22903-2475 (United States)] [National Radio Astronomy Observatory, 520 Edgemont Road, Charlottesville, VA, 22903-2475 (United States); Anderson, L. D. [Department of Physics, West Virginia University, Morgantown, WV 26506 (United States)] [Department of Physics, West Virginia University, Morgantown, WV 26506 (United States)

2013-02-10T23:59:59.000Z

328

Theory and modelling of helium enrichment in plasma experiments with pump limiters  

DOE Green Energy (OSTI)

Helium enrichment in the exhaust gas stream flowing from a hydrogen-helium plasma is studied using an analytical theory and Monte Carlo simulations. To provide a sensitive experimental test in a tokamak, an unusual configuration, inverted from traditional designs, is proposed for a pump limiter. The principle can be tested in other plasma devices as well. The theory suggests that for typical plasma edge conditions in a confinement device, namely, n = 10/sup 13/cm/sup -3/ and T/sub i/ = T/sub e/ approx. = 5-30eV, helium enrichment in the neutral gas exhaust stream can be very high, in the range 5 to 7, relative to the helium-hydrogen ratio in the plasma. Such high enrichment factors are achieved by exploiting the difference between the ionization rates of hydrogen and helium and the negligible helium charge exchange rate at these plasma conditions. A limiter arrangement is proposed in which the natural curvature of the toroidal magnetic field is used to isolate, using the plasma itself, the point of plasma neutralization from the location of the gas exhaust. The plasma region then acts to preferentially screen the recycling hydrogen by the processes of ionization and of charge-exchange-induced losses at open boundaries. The theory and analysis suggests that an experiment can provide a sensitive test of modules used to describe the plasma edge and of atomic and surface physics data used in these models.

Prinja, A.K.; Conn, R.W.

1984-01-01T23:59:59.000Z

329

NEUTRONIC REACTOR CONSTRUCTION AND OPERATION  

DOE Patents (OSTI)

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.

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

1960-03-15T23:59:59.000Z

330

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

Science Conference Proceedings (OSTI)

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 employed. It also involves validating the computer code and turbulence model against a series of separate and combined flow phenomena and selecting the data used for the validation. This report describes progress made to identify proper modeling strategies for simulating the lower plenum flow for the task entitled “CFD software validation of jets in crossflow,” which was designed to investigate the issues pertaining to the validation process. The flow phenomenon previously chosen to investigate is flow in a staggered tube bank because it is shown by preliminary simulations to be the location of the highest turbulence intensity in the lower plenum Numerical simulations were previously obtained assuming that the flow is steady. Various turbulence models were employed along with strategies to reduce numerical error to allow appropriate comparisons of the results. It was determined that the sophisticated Reynolds stress model (RSM) provided the best results. It was later determined that the flow is an unsteady flow wherein circulating eddies grow behind the tube and ‘peel off’ alternately from the top and the bottom of the tube. Additional calculations show that the mean velocity is well predicted when the flow is modeled as an unsteady flow. The results for U are clearly superior for the unsteady computations; the unsteady computations for the turbulence stress are similar to those for the steady calculations, showing the same trends. It is clear that strategie

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

2006-09-01T23:59:59.000Z

331

Conditions For Successful Helium Detonations In Astrophysical Environments  

E-Print Network (OSTI)

Several models for type Ia-like supernovae events rely on the production of a self-sustained detonation powered by nuclear reactions.In the absence of hydrogen, the fuel that powers these detonations typically consists of either pure helium (He) or a mixture of carbon and oxygen (C/O). Studies that systematically determine the conditions required to initiate detonations in C/O material exist, but until now no analogous investigation of degenerate He matter has been conducted. We perform one-dimensional reactive hydrodynamical simulations at a variety of initial density and temperature combinations and find critical length scales for the initiation of He detonations that range between 1 -- $10^{10}$ cm. These sizes are consistently smaller than the corresponding Chapman-Jouguet (CJ) length scales by a factor of ~100, providing opportunities for thermonuclear explosions in a wider range of low mass white dwarfs (WDs) than previously thought possible. We find that virialized WDs with as little mass as 0.24 $M_\\o...

Holcomb, Cole; De Colle, Fabio; Ramirez-Ruiz, Enrico

2013-01-01T23:59:59.000Z

332

Helium Behavior in Oxide Nuclear Fuels: First Principles Modeling  

Science Conference Proceedings (OSTI)

UO2 and (U, Pu)O2 solid solutions (the so-called MOX) nowadays are used as commercial nuclear fuels in many countries. One of the safety issues during the storage of these fuels is related to their self-irradiation that produces and accumulates point defects and helium therein. We present density functional theory (DFT) calculations for UO2, PuO2 and MOX containing He atoms in octahedral interstitial positions. In particular, we calculated basic MOX properties and He incorporation energies as functions of Pu concentration within the spin-polarized, generalized gradient approximation (GGA) DFT calculations. We also included the on-site electron correlation corrections using the Hubbard model (in the framework of the so-called DFT + U approach). We found that PuO2 remains semiconducting with He in the octahedral position while UO2 requires a specific lattice distortion. Both materials reveal a positive energy for He incorporation, which, therefore, is an exothermic process. The He incorporation energy increases with the Pu concentration in the MOX fuel.

Gryaznov, D.; Rashkeev, Sergey N.; Kotomin, E. A.; Heifets, Eugene; Zhukovskii, Yuri F.

2010-10-01T23:59:59.000Z

333

Once-ionized helium in superstrong magnetic fields  

E-Print Network (OSTI)

It is generally believed that magnetic fields of some neutron stars, the so-called magnetars, are enormously strong, up to 10^{14} - 10^{15} G. Recent investigations have shown that the atmospheres of magnetars are possibly composed of helium. We calculate the structure and bound-bound radiative transitions of the He^+ ion in superstrong fields, including the effects caused by the coupling of the ion's internal degrees of freedom to its center-of-mass motion. We show that He^+ in superstrong magnetic fields can produce spectral lines with energies of up to about 3 keV, and it may be responsible for absorption features detected recently in the soft X-ray spectra of several radio-quiet isolated neutron stars. Quantization of the ion's motion across a magnetic field results in a fine structure of spectral lines, with a typical spacing of tens electron-volts in magnetar-scale fields. It also gives rise to ion cyclotron transitions, whose energies and oscillator strengths depend on the state of the bound ion.

George G. Pavlov; Victor G. Bezchastnov

2005-05-23T23:59:59.000Z

334

Methane coupling by membrane reactor. Quarterly technical progress report, December 25, 1995--March 24, 1996  

DOE Green Energy (OSTI)

Methane oxidative coupling experiments were conducted in a packed-bed reactor using La/MgO catalyst. The results were in agreement with those reported in the literature. Oxygen permeability through an oxygen-conductive dense membrane was measured. The oxygen permeability was about 20% of the reported values and increased with the sweep flow rate of helium in the tube side. Blank runs under methane coupling conditions were carried out with no catalyst packed in the dense membrane tube. Nearly 100% CO{sub 2} selectivity was observed, suggesting that the membrane material is a facile total oxidation catalyst. After the experiment the membrane tube was analyzed by scanning electron microscopy.

NONE

1996-05-28T23:59:59.000Z

335

Refrigeration requirements for fusion reactors based upon the theta-pinch concept  

DOE Green Energy (OSTI)

Two refrigeration systems applicable to the theta-pinch fusion concept are described. The first is a 1100 W, 4.5 K refrigerator which will be used for testing superconducting NbTi Magnetic Energy Transfer and Storage (METS) coil systems. This unit is currently being installed and is to be operational by April 1977. The second unit is applicable to the Syllac Fusion Test Reactor (SFTR) and has been conceptually designed. This liquefier-refrigerator is about 22 times larger than those in existence at present and will require 12-MW input electrical power. It will provide 3045 kg/h of liquid helium at 4.5 K.

Williamson, K.D. Jr.; King, C.R.

1976-01-01T23:59:59.000Z

336

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

DOE Green Energy (OSTI)

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.

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

2008-08-01T23:59:59.000Z

337

THE HGCR-1, A DESIGN STUDY OF A NUCLEAR POWER STATION EMPLOYING A HIGH- TEMPERATURE GAS-COOLED REACTOR WITH GRAPHITE-UO$sub 2$ FUEL ELEMENTS  

SciTech Connect

The preliminary design of a 3095-Mw(thermal), helium-cooled, graphite- moderated reactor employing sign conditions, 1500 deg F reactor outlet gas would be circulated to eight steam generators to produce 1050 deg F, 1450-psi steam which would be converted to electrical power in eight 157-Mw(electrical) turbine- generators. The over-all efficiency of this nuclear power station is 36.5%. The significant activities released from the unclad graphite-UO/sub 2/ fuel appear to be less than 0.2% of those produced and would be equivalent to 0.002 curie/ cm/ sup 3/ in the primary helium circuit. The maintenance problems associated with this contamination level are discussed. A cost analysis indicates that the capital cost of this nuclear station per electrical kilowatt would be around 0, and that the production cost of electrical power would be 7.8 mills/kwhr. (auth)

Cottrell, W.B.; Copenhaver, C.M.; Culver, H.N.; Fontana, M.H.; Kelleghan, V.J.; Samuels, G.

1959-07-28T23:59:59.000Z

338

REACTOR GROUT THERMAL PROPERTIES  

DOE Green Energy (OSTI)

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.

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

2011-01-28T23:59:59.000Z

339

Hartree-Fock calculations for the photoionisation of helium and helium-like ions in neutron star magnetic fields  

E-Print Network (OSTI)

We derive the photoionisation cross section in dipole approximation for many-electron atoms and ions for neutron star magnetic field strengths in the range of 10^7 to 10^9 T. Both bound and continuum states are treated in adiabatic approximation in a self-consistent way. Bound states are calculated by solving the Hartree-Fock-Roothaan equations using finite-element and B-spline techniques while the continuum orbital is calculated by direct integration of the Hartree-Fock equations in the mean-field potential of the remaining bound orbitals. We take into account mass and photon density in the neutron star's atmosphere as well as thermal occupation of the levels. The data are of importance for the quantitative interpretation of observed x-ray spectra that originate from the thermal emission of isolated neutron stars. They can serve as input for modeling neutron star atmospheres as regards chemical composition, magnetic field strength, temperature, and redshift. Our main focus in this paper lies on helium, heliu...

Kersting, Thorsten; Diemand, Peter; Wunner, Günter

2013-01-01T23:59:59.000Z

340

B Reactor | Department of Energy  

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

Operational Management » History » Manhattan Project » Signature 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 large-scale plutonium production reactor. As at Oak Ridge, the need for labor turned Hanford into an atomic 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 on a much larger scale and used water rather than air as a coolant. Whereas the X-10 had an initial design output of 1,000 kilowatts, the B Reactor was designed to operate at 250,000 kilowatts. Consisting of a 28- by 36-foot, 1,200-ton graphite cylinder lying on its side, the reactor was penetrated through its

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341

Zero Power Reactor simulation | Argonne National Laboratory  

NLE Websites -- All DOE Office Websites (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...

342

THE MATERIALS OF FAST BREEDER REACTORS  

E-Print Network (OSTI)

jet aircraft engines, and nuclear reactor fuel elements. Ancomponents of a nuclear reactor core are susceptible tothe nuclear physics of the thermal and fast neutron reactors

Olander, Donald R.

2013-01-01T23:59:59.000Z

343

Advanced Reactor Development and Technology - Nuclear Engineering...  

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

Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Nuclear Data Program Advanced Reactor Development Overview Advanced Fast Reactor...

344

D0 Silicon Upgrade: Commissioning Test Results for D-Zero's Helium Refrigerator  

Science Conference Proceedings (OSTI)

The test objectives are: (1) Make liquid helium and measure refrigerator capacity; (2) Measure liquid helium dewar heat leak, transfer line heat leak, and liquid nitrogen consumption rates; (3) Operate all cryogenic transfer lines; (4) Get some running time on all components; (5) Debug mechanical components, instrumentation, DMACs user interface, tune loops, and otherwise shake out any problems; (6) Get some operating time in to get familiar with system behavior; (7) Revise and/or improve operating procedures to actual practice; and (8) Identify areas for future improvement. D-Zero's stand alone helium refrigerator (STAR) liquified helium at a rate of 114 L/hr. This is consistent with other STAR installations. Refrigeration capacity was not measured due to lack of a calibrated heat load. Measured heat leaks were within design values. The helium dewar loss was measured at 2 to 4 watts or 9% per day, the solenoid and VLPC helium transfer lines had a heat leak of about 20 watts each. The liquid nitrogen consumption rates of the mobile purifier, STAR, and LN2 subcooler were measured at 20 gph, 20 to 64 gph, and 3 gph respectively. All cryogenic transfer lines including the solenoid and visible light photon counter (VLPC) transfer lines were cooled to their cryogenic operating temperatures. This included independent cooling of nitrogen shields and liquid helium components. No major problems were observed. The system ran quite well. Many problems were identified and corrected as they came up. Areas for improvement were noted and will be implemented in the future. The instrumentation and control system operated commendably during the test. The commissioning test run was a worthwhile and successful venture.

Rucinski, Russ; /Fermilab

1997-06-30T23:59:59.000Z

345

ALTERNATIVES TO HELIUM-3 FOR NEUTRON MULTIPLICITY DETECTORS  

Science Conference Proceedings (OSTI)

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

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

2012-02-07T23:59:59.000Z

346

Hydrogen and helium traces in type Ib-c supernovae  

E-Print Network (OSTI)

The spectroscopic properties of a selected optical photospheric spectra of core collapse supernovae (CCSNe) are investigated.Special attention is devoted to traces of hydrogen at early phases. The generated spectra are found to match the observed ones reasonably well, including a list of only 23 candidate ions. Guided by SN Ib 1990I, the observed trough near 6300\\AA is attributed to H$\\alpha$ in almost all Type Ib events, although in some objects it becomes too weak to be discernible, especially at later phases. Alternative line identifications are discussed. Differences in the way hydrogen manifests its presence within CCSNe are highlighted. In Type Ib SNe, the H$\\alpha$ contrast velocity (i.e. line velocity minus the photospheric velocity) seems to increase with time at early epochs, reaching values as high as 8000 km s$^{-1}$ around 15-20 days after maximum and then remains almost constant. The derived photospheric velocities, indicate a lower velocity for Type II SNe 1987A and 1999em as compared to SN Ic 1994I and SN IIb 1993J, while Type Ib events display a somewhat larger variation. The scatter, around day 20, is measured to be $\\sim$5000 km s$^{-1}$. Following two simple approaches, rough estimates of ejecta and hydrogen masses are given. A mass of hydrogen of approximately 0.02 $M_\\odot$ is obtained for SN 1990I, while SNe 1983N and 2000H ejected $\\sim$0.008 $M_\\odot$ and $\\sim$0.08 $M_\\odot$ of hydrogen, respectively. SN 1993J has a higher hydrogen mass, $\\sim 0.7$ $M_\\odot$ with a large uncertainty. A low mass and thin hydrogen layer with very high ejection velocities above the helium shell, is thus the most likely scenario for Type Ib SNe. Some interesting and curious issues relating to oxygen lines suggest future investigations.

A. Elmhamdi; I. J. Danziger; D. Branch; B. Leibundgut; E. Baron; R. P. Kirshner

2005-12-22T23:59:59.000Z

347

Nuclear reactor safety device  

DOE Patents (OSTI)

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.

Hutter, Ernest (Wilmette, IL)

1986-01-01T23:59:59.000Z

348

REACTOR CONTROL DEVICE  

DOE Patents (OSTI)

A wholly mechanical compact control device is designed for automatically rendering the core of a fission reactor subcritical in response to core temperatures in excess of the design operating temperature limit. The control device comprises an expansible bellows interposed between the base of a channel in a reactor core and the inner end of a fuel cylinder therein which is normally resiliently urged inwardly. The bellows contains a working fluid which undergoes a liquid to vapor phase change at a temperature substantially equal to the design temperature limit. Hence, the bellows abruptiy expands at this limiting temperature to force the fuel cylinder outward and render the core subcritical. The control device is particularly applicable to aircraft propulsion reactor service. (AEC)

Graham, R.H.

1962-09-01T23:59:59.000Z

349

A NEUTRONIC REACTOR  

DOE Patents (OSTI)

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.

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

1959-09-01T23:59:59.000Z

350

MERCHANT MARINE SHIP REACTOR  

DOE Patents (OSTI)

A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

1961-05-01T23:59:59.000Z

351

Heat dissipating nuclear reactor  

DOE Patents (OSTI)

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.

Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

1987-01-01T23:59:59.000Z

352

Heat dissipating nuclear reactor  

DOE Patents (OSTI)

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.

Hunsbedt, A.; Lazarus, J.D.

1985-11-21T23:59:59.000Z

353

Reactor for exothermic reactions  

DOE Patents (OSTI)

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.

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

1993-03-02T23:59:59.000Z

354

MERCHANT MARINE SHIP REACTOR  

DOE Patents (OSTI)

A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

Sankovich, M.F.; Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Gestson, D.K.

1961-05-01T23:59:59.000Z

355

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from

356

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations, which govern the operation of commercial nuclear power plants, require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including

357

NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) PROGRAM GRANT NUMBER DE-FG03-00SF22168 TECHNICAL PROGRESS REPORT (Nov. 15, 2001 - Feb. 15,2002) ''Design and Layout Concepts for Compact, Factory-Produced, Transportable, Generation IV Reactor Systems''  

SciTech Connect

The objectives of this project are to develop and evaluate nuclear power plant designs and layout concepts to maximize the benefits of compact modular Generation IV reactor concepts including factory fabrication and packaging for optimal transportation and siting. Three nuclear power plant concepts are being studied representing water, helium and lead-bismuth coolants. This is the sixth quarterly progress report.

Fred R. Mynatt; Andy Kadak; Marc Berte; Larry Miller; Mohammed Khan; Joe McConn; Lawrence Townsend; Wesley Williams; Martin Williamson

2002-03-15T23:59:59.000Z

358

Nuclear reactor apparatus  

DOE Patents (OSTI)

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.

Wade, Elman E. (Ruffs Dale, PA)

1978-01-01T23:59:59.000Z

359

Fusion reactor pumped laser  

DOE Patents (OSTI)

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.

Jassby, Daniel L. (Princeton, NJ)

1988-01-01T23:59:59.000Z

360

Perspectives on reactor safety  

SciTech Connect

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.

Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

1994-03-01T23:59:59.000Z

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


361

Reactor Neutrino Experiments  

E-Print Network (OSTI)

Precisely measuring $\\theta_{13}$ is one of the highest priority in neutrino oscillation study. Reactor experiments can cleanly determine $\\theta_{13}$. Past reactor neutrino experiments are reviewed and status of next precision $\\theta_{13}$ experiments are presented. Daya Bay is designed to measure $\\sin^22\\theta_{13}$ to better than 0.01 and Double Chooz and RENO are designed to measure it to 0.02-0.03. All are heading to full operation in 2010. Recent improvements in neutrino moment measurement are also briefed.

Jun Cao

2007-12-06T23:59:59.000Z

362

Diagnostics for hybrid reactors  

SciTech Connect

The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

Orsitto, Francesco Paolo [ENEA Unita' Tecnica Fusione , Associazione ENEA-EURATOM sulla Fusione C R Frascati v E Fermi 45 00044 Frascati (Italy)

2012-06-19T23:59:59.000Z

363

THERMAL NUCLEAR REACTOR  

DOE Patents (OSTI)

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.

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

1957-09-24T23:59:59.000Z

364

MEANS FOR SHIELDING REACTORS  

DOE Patents (OSTI)

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.

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

1959-03-10T23:59:59.000Z

365

Fast quench reactor method  

DOE Patents (OSTI)

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.

Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID); Berry, Ray A. (Idaho Falls, ID)

1999-01-01T23:59:59.000Z

366

Reactor sharing experience at the MIT research reactor  

SciTech Connect

This paper provides a number of examples of how educational institutions in the Boston area and elsewhere that do not possess nuclear reactors for training and research purposes have successfully enriched their programs through utilization of the Massachusetts Institute of Technology Research Reactor (MITR) with assistance from the Reactor Sharing Program of the US Department of Energy (DOE).

Clark, L. Jr.; Fecych, W.; Young, H.H.

1985-01-01T23:59:59.000Z

367

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

SciTech Connect

During FY 2011 the DB Program will report Highlights on a monthly basis, but will no longer produce Quarterly Progress Reports. Technical details that were previously included in the quarterly reports will be included in the appropriate Milestone Reports that are submitted to FCRD Program Management. These reports will also be uploaded to the Deep Burn website. The Monthly Highlights report for October 2010, ORNL/TM-2010/300, was distributed to program participants on November 29, 2010. This report discusses the following: (1) Thermochemical Data and Model Development; (2) TRU (transuranic elements) TRISO (tri-structural isotropic) Development - (a) TRU Kernel Development, (b) Coating Development; (3) LWR Fully Ceramic Fuel - (a) FCM Fabrication Development, (b) FCM Irradiation Testing.

Snead, Lance Lewis [ORNL; Bell, Gary L [ORNL; Besmann, Theodore M [ORNL

2010-12-01T23:59:59.000Z

368

Estimation of tritium and helium inventory in the tritium handling system in Korea  

Science Conference Proceedings (OSTI)

In Korea, the Wolsong Tritium Removal Facility (WTRF) is under construction to reduce the amount of tritium present in the moderator and coolant of the CANDU type Wolsong nuclear power plants. Recently, a study on the tritium handling system for recovery of the tritium collected from the WTRF was started. Some tritium would enter the steel of the container walls and subsequently decay to helium. This helium can deteriorate the mechanical properties of the material of the tritium handling system. To evaluate the tritium and helium inventory in the stainless steel wall of this system, the time-dependent diffusion equation was developed, solved and the results are presented in this paper. These results were compared to previous work that evaluated the tritium inventory in the stainless steel wall of 50-L tritium containers. Tritium and helium concentration profiles and the corresponding inventories were evaluated with respect to the various parameters such as exposure time, temperature, and partial pressure. After 24 years, the helium inventory in the wall of the tritium handling system exceeds the tritium inventory. (authors)

Yook, D.; Lee, S.; Lee, K. [Dept. of Nuclear Eng., KAIST, 373-1, Kusong-dong, Yusong-gu, Daejon 305-701 (Korea, Republic of); Song, K. M.; Shon, S. H. [KEPRI, 103-16 Munji-Dong, Yuseong-Gu, Daejeon, 305-380 (Korea, Republic of)

2008-07-15T23:59:59.000Z

369

Terracentric Nuclear Fission Reactor: Background, Basis, Feasibility, Structure, Evidence, and Geophysical Implications  

E-Print Network (OSTI)

The background, basis, feasibility, structure, evidence, and geophysical implications of a naturally occurring Terracentric nuclear fission georeactor are reviewed. For a nuclear fission reactor to exist at the center of the Earth, all of the following conditions must be met: (1) There must originally have been a substantial quantity of uranium within Earth's core; (2) There must be a natural mechanism for concentrating the uranium; (3) The isotopic composition of the uranium at the onset of fission must be appropriate to sustain a nuclear fission chain reaction; (4) The reactor must be able to breed a sufficient quantity of fissile nuclides to permit operation over the lifetime of Earth to the present; (5) There must be a natural mechanism for the removal of fission products; (6) There must be a natural mechanism for removing heat from the reactor; (7) There must be a natural mechanism to regulate reactor power level, and; (8) The location of the reactor or must be such as to provide containment and prevent meltdown. Herndon's georeactor alone is shown to meet those conditions. Georeactor existence evidence based upon helium measurements and upon antineutrino measurements is described. Geophysical implications discussed include georeactor origin of the geomagnetic field, geomagnetic reversals from intense solar outbursts and severe Earth trauma, as well as georeactor heat contributions to global dynamics.

J. Marvin Herndon

2013-08-23T23:59:59.000Z

370

Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm  

SciTech Connect

There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

Drajat, R. Z.; Su'ud, Z.; Soewono, E.; Gunawan, A. Y. [Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Physics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia)

2012-05-22T23:59:59.000Z

371

THE HOMOGENEOUS SUSPENSION REACTOR PROJECT  

SciTech Connect

The considerations which led to the study of a homogeneous suspension reactor are reviewed briefly. The characteristics of the KEMA Suspension Test Reactor (KSTR) are then summarized. (J.S.R.)

Went, J.J.

1963-02-01T23:59:59.000Z

372

Advanced Nuclear Reactors | Department of Energy  

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

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...

373

Innovative design of uranium startup fast reactors  

E-Print Network (OSTI)

Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

Fei, Tingzhou

2012-01-01T23:59:59.000Z

374

Reactor operation safety information document  

Science Conference Proceedings (OSTI)

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)

Not Available

1990-01-01T23:59:59.000Z

375

NEUTRONIC REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A reactor fuel element of the capillary tube type is described. The element consists of a thin walled tube, sealed at both ends, and having an interior coatlng of a fissionable material, such as uranium enriched in U-235. The tube wall is gas tight and is constructed of titanium, zirconium, or molybdenum.

Kesselring, K.A.; Seybolt, A.U.

1958-12-01T23:59:59.000Z

376

NEUTRONIC REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A fuel slug for a reactor which acts as a safety device is described. The fuel slug is an aluminum tube with a foil lining the inside surface of the tube, the foil being fabricated of uranium in a lead matrix.

Horning, W.A.; Lanning, D.D.; Donahue, D.J.

1959-10-01T23:59:59.000Z

377

WATER BOILER REACTOR  

DOE Patents (OSTI)

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

King, L.D.P.

1960-11-22T23:59:59.000Z

378

The First Reactor  

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

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 Alonzo Stagg Field stadium, at the University of...

379

Thermal Reactor Safety  

Science Conference Proceedings (OSTI)

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.

Not Available

1980-06-01T23:59:59.000Z

380

NEUTRONIC REACTOR FUEL PUMP  

DOE Patents (OSTI)

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.)

Cobb, W.G.

1959-06-01T23:59:59.000Z

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


381

Cermet fuel reactors  

Science Conference Proceedings (OSTI)

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.

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

1987-09-01T23:59:59.000Z

382

Nuclear reactor building  

DOE Patents (OSTI)

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.

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

1994-04-05T23:59:59.000Z

383

Fossil fuel furnace reactor  

DOE Patents (OSTI)

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.

Parkinson, William J. (Los Alamos, NM)

1987-01-01T23:59:59.000Z

384

Nuclear Reactors and Technology  

SciTech Connect

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.

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

1992-01-01T23:59:59.000Z

385

Nuclear reactor building  

DOE Patents (OSTI)

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.

Gou, Perng-Fei (Saratoga, CA); Townsend, Harold E. (Campbell, CA); Barbanti, Giancarlo (Sirtori, IT)

1994-01-01T23:59:59.000Z

386

Reactor component automatic grapple  

DOE Patents (OSTI)

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.

Greenaway, Paul R. (Bethel Park, PA)

1982-01-01T23:59:59.000Z

387

NUCLEAR REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

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

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

1963-06-11T23:59:59.000Z

388

NEUTRONIC REACTOR STRUCTURE  

DOE Patents (OSTI)

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.

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

1961-05-30T23:59:59.000Z

389

NEUTRONIC REACTOR STRUCTURE  

DOE Patents (OSTI)

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)

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

1961-05-30T23:59:59.000Z

390

NEUTRONIC REACTOR SHIELD  

DOE Patents (OSTI)

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.

Fermi, E.; Zinn, W.H.

1957-09-24T23:59:59.000Z

391

NEUTRONIC REACTOR CONTROL ELEMENT  

DOE Patents (OSTI)

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)

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

1962-04-17T23:59:59.000Z

392

Progress on the conceptual design of a mirror hybrid fusion--fission reactor  

SciTech Connect

A conceptual design study was made of a fusion-fission reactor for the purpose of producing fissile material and electricity. The fusion component is a D-T plasma confined by a pair of magnetic mirror coils in a Yin-Yang configuration and is sustained by neutral beam injection. The neutrons from the fusion plasma drive the fission assembly which is composed of natural uranium carbide fuel rods clad with stainless steel and helium cooled. It was shown conceptually how the reactor might be built using essentially present-day technology and how the uranium-bearing blanket modules can be routinely changed to allow separation of the bred fissile fuel. (MOW)

Moir, R.W.; Lee, J.D.; Burleigh, R.J.

1975-06-25T23:59:59.000Z

393

HELIUM SHELL DETONATIONS ON LOW-MASS WHITE DWARFS AS A POSSIBLE EXPLANATION FOR SN 2005E  

SciTech Connect

Recently, several Type Ib supernovae (SNe; with the prototypical SN 2005E) have been shown to have atypical properties. These SNe are faint (absolute peak magnitude of {approx} - 15) and fast SNe that show unique composition. They are inferred to have low ejecta mass (a few tenths of a solar mass) and to be highly enriched in calcium, but poor in silicon elements and nickel. These SNe were therefore suggested to belong to a new class of calcium-rich faint SNe explosions. Their properties were proposed to be the result of helium detonations that may occur on helium accreting white dwarfs. In this paper, we theoretically study the scenario of helium detonations and focus on the results of detonations in accreted helium layers on low-mass carbon-oxygen (CO) cores. We present new results from one-dimensional simulations of such explosions, including their light curves and spectra. We find that when the density of the helium layer is low enough the helium detonation produces large amounts of intermediate elements, such as calcium and titanium, together with a large amount of unburnt helium. Alternatively, enough carbon enrichment of the accreted helium as a result of convective undershoot at the early stages of the runaway can avoid the production of iron group elements as the alpha particles are consumed avoiding iron production. Our results suggest that the properties of calcium-rich faint SNe could indeed be consistent with the helium-detonation scenario on small CO cores. Above a certain density (larger CO cores) the detonation leaves mainly {sup 56}Ni and unburnt helium, and the predicted spectrum will unlikely fit the unique features of this class of SNe. Finally, none of our studied models reproduces the bright, fast-evolving light curves of another type of peculiar SNe suggested to originate in helium detonations (SNe 1885A, 1939B, and 2002bj).

Waldman, Roni; Livne, Eli; Glasner, Ami [Racah Institute of Physics, Hebrew University, Jerusalem 91904 (Israel); Sauer, Daniel [Stockholm University, Department for Astronomy, AlbaNova University Center, 106 91 Stockholm (Sweden); Perets, Hagai [Harvard-Smithsonian Center for Astrophysics, 60 Garden Street, Cambridge, MA 02138 (United States); Mazzali, Paolo [Max-Planck Institut fuer Astrophysik, Karl-Schwarzschild-Str. 1, 85748 Garching (Germany); Truran, James W. [Physics Division, Argonne National Laboratory, Argonne, IL 60439 (United States); Gal-Yam, Avishay [Benoziyo Center for Astrophysics, Faculty of Physics, Weizmann Institute of Science, Rehovot 76100 (Israel)

2011-09-01T23:59:59.000Z

394

REACTOR FUEL ELEMENTS TESTING CONTAINER  

DOE Patents (OSTI)

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)

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

1963-01-15T23:59:59.000Z

395

EMERGENCY SHUTDOWN FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

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)

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

1963-12-24T23:59:59.000Z

396

Modeling Chemical Reactors I: Quiescent Reactors  

E-Print Network (OSTI)

We introduce a fully generalized quiescent chemical reactor system in arbitrary space $\\vdim =1,2$ or 3, with $n\\in\\mathbb{N}$ chemical constituents $\\alpha_{i}$, where the character of the numerical solution is strongly determined by the relative scaling between the local reactivity of species $\\alpha_{i}$ and the local functional diffusivity $\\mathscr{D}_{ij}(\\alpha)$ of the reaction mixture. We develop an operator time-splitting predictor multi-corrector RK--LDG scheme, and utilize $hp$-adaptivity relying only on the entropy $\\mathscr{S}_{\\mathfrak{R}}$ of the reactive system $\\mathfrak{R}$. This condition preserves these bounded nonlinear entropy functionals as a necessarily enforced stability condition on the coupled system. We apply this scheme to a number of application problems in chemical kinetics; including a difficult classical problem arising in nonequilibrium thermodynamics known as the Belousov-Zhabotinskii reaction where we utilize a concentration-dependent diffusivity tensor $\\mathscr{D}_{ij}(...

Michoski, C E; Schmitz, P G

2010-01-01T23:59:59.000Z

397

Flow Of Mantle Fluids Through The Ductile Lower Crust- Helium Isotope  

Open Energy Info (EERE)

Of Mantle Fluids Through The Ductile Lower Crust- Helium Isotope Of Mantle Fluids Through The Ductile Lower Crust- Helium Isotope Trends Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: Flow Of Mantle Fluids Through The Ductile Lower Crust- Helium Isotope Trends Details Activities (5) Areas (5) Regions (0) Abstract: Heat and mass are injected into the shallow crust when mantle fluids are able to flow through the ductile lower crust. Minimum He-3/He-4 ratios in surface fluids from the northern Basin and Range Province, western North America, increase systematically from low crustal values in the east to high mantle values in the west, a regional trend that correlates with the rates of active crustal deformation. The highest ratios occur where the extension and shear strain rates are greatest. The

398

Cooling Strings of Superconducting Devices below 2 K the Helium II Bayonet Heat Exchanger  

E-Print Network (OSTI)

High-energy particle accelerators and colliders contain long strings of superconducting devices - acceleration RF cavities and magnets - operating at high field, which may require cooling in helium II below 2 K. In order to maintain adequate operating conditions, the applied or generated heat loads must be extracted and transported with minimum temperature difference. Conventional cooling schemes based on conductive or convective heat transport in pressurized helium II very soon reach their intrinsic limits of thermal impedance over extended lengths. We present the concept of helium II bayonet heat exchanger, which has been developed at CERN for the magnet cooling scheme of the Large Hadron Collider (LHC), and describe its specific advantages as a slim, quasi-isothermal heat sink. Experimental results obtained on several test set-ups, and a prototype magnet string have permitted to validate its performance and sizing rules, for transporting linear heat loads in the W.m-1 range over distances of several tens o...

Lebrun, P; Tavian, L; Van Weelderen, R

1998-01-01T23:59:59.000Z

399

Survey of Optimization of Reactor Coolant Cleanup Systems: For Boiling Water Reactors and Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Optimization of the reactor coolant cleanup systems in the boiling water reactor (BWR) and pressurized water reactor (PWR) environment is important for controlling the transport of corrosion products (metals and activated metals), fission products, and coolant impurities (soluble and insoluble) throughout the reactor coolant loop, and this optimization contributes to reducing primary system radiation fields. The removal of radionuclides and corrosion products is just one of many functions (both ...

2013-09-27T23:59:59.000Z

400

Helium and Hydrogen Release Measurements on Various Alloys Irradiated in SINQ  

SciTech Connect

Three irradiations have been performed in the Swiss Spallation Neutron Source (SINQ) to establish a materials database for mixed proton and neutron fluxes for future spallation neutron and other accelerator sources. Materials included in the second irradiation, STIP-II, included mainly austenitic and martensitic steels. Samples of 316LN, F82H, Al, and Zircaloy-2 from STIP ?II have been analyzed for their total helium and hydrogen contents and their release characteristics. In terms of total gas content, the helium and hydrogen results are similar to those observed earlier from STIP-I. Specifically, the helium contents tended to be somewhat higher (up to 50%) than calculated for both low and high dose samples. 3He/4He ratios were generally in agreement with expectations except for the Ziracaloy-2 which showed very low values, likely from increased 3He generation from decay of irradiation-generated tritium. Hydrogen contents on the other hand tended inversely with dose, suggesting increased hydrogen loss at the higher dose (and higher temperature) locations from diffusion. Hydrogen levels in the Zircaloy were considerably higher than expected ({approx}10,000 to 25,000 appm), suggesting additional pickup of hydrogen from the irradiation environment. Hydrogen levels in the aluminum were similar to predicted. Helium and hydrogen release measurements from the temperature ramp experiments showed considerable levels of deuterium and tritium species which generally mirrored those of hydrogen. Hydrogen release occurred from about 300 for the aluminum to about 800 C for the Zircaloy-2. For the Zircaloy-2 and the steels, helium release began to occur at about 1100 C, which is consistent with previous measurements on irradiated steels. For the aluminum, helium release began at just under 600 C.

Oliver, Brian M.; Dai, Yong; Causey, Rion A.

2006-09-15T23:59:59.000Z

Note: This page contains sample records for the topic "helium reactor gt-mhr" from the National Library of EnergyBeta (NLEBeta).
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to obtain the most current and comprehensive results.


401

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

SciTech Connect

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

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

1995-04-01T23:59:59.000Z

402

Reactor vessel support system. [LMFBR  

DOE Patents (OSTI)

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.

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

1980-05-09T23:59:59.000Z

403

Solutions for Liquid Nitrogen Pre-Cooling in Helium Refrigeration Cycles  

E-Print Network (OSTI)

Pre-cooling of helium by means of liquid nitrogen is the oldest and one of the most common process features used in helium liquefiers and refrigerators. Its two principle tasks are to allow or increase the rate of pure liquefaction, and to permit the initial cool-down of large masses to about 80 K. Several arrangements for the pre-cooling process are possible depending on the desired application. Each arrangement has its proper advantages and drawbacks. The aim of this paper is to review the possible process solutions for liquid nitrogen pre-cooling and their particularities.

Wagner, U

2000-01-01T23:59:59.000Z

404

Detection and Imaging of He{sub 2} Molecules in Superfluid Helium  

SciTech Connect

We present data that show a cycling transition can be used to detect and image metastable He{sub 2} triplet molecules in superfluid helium. We demonstrate that limitations on the cycling efficiency due to the vibrational structure of the molecule can be mitigated by the use of repumping lasers. Images of the molecules obtained using the method are also shown. This technique gives rise to a new kind of ionizing radiation detector. The use of He{sub 2} triplet molecules as tracer particles in the superfluid promises to be a powerful tool for visualization of both quantum and classical turbulence in liquid helium.

Rellergert, W. G.; Cahn, S. B.; Garvan, A.; Hanson, J. C.; Lippincott, W. H.; Nikkel, J. A.; McKinsey, D. N. [Department of Physics, Yale University, New Haven, Connecticut 06520 (United States)

2008-01-18T23:59:59.000Z

405

Particle Detection in Superfluid Helium: R&D for Low Energy Solar Neutrinos  

DOE Green Energy (OSTI)

This report presents a summary of the results from R&D conducted as a feasibility study in the Department of Physics of Brown University for detection of low energy solar neutrinos utilizing a superfluid helium target. The report outlines the results in several areas: 1) development of experimental facilities, 2) energy deposition by electrons and alphas in superfluid helium, 3) development of wafer and metallic magnetic calorimeters, 4) background studies, 5) coded apertures and conceptual design, 6) Detection of single electrons and 7) a simulation of expected performance of a full scale device. Recommendations for possible future work are also presented. A bibliography of published papers and unpublished doctoral theses is included.

Lanou, Robert E., Jr.

2006-03-31T23:59:59.000Z

406

Nanometer-scale tunnel formation in metallic glass by helium ion irradiation  

SciTech Connect

We have shown that upon high fluence helium ion irradiation, metallic glass Cu{sub 50}Zr{sub 45}Ti{sub 5} becomes highly porous at the depth of the helium projected range. The resulting porous region is characterized by the formation of a tunnel like structure and self-linkage of nanometer size gas bubbles. Furthermore, the irradiation leads to the formation of nanometer size Cu{sub x}Zr{sub y} crystals that are randomly distributed. The results of this study indicate that the He-filled bubbles have attractive interactions and experience considerable mobility. Movement of the bubbles is believed to be assisted by ballistic collisions.

Shao Lin [Department of Nuclear Engineering, Texas A and M University, College Station, Texas 77843 (United States); Materials Science and Engineering Program, Texas A and M University, College Station, Texas 77843 (United States); Gorman, Brian P. [Department of Metallurgical and Materials Engineering, Colorado School of Mines, Golden, Colorado 80401 (United States); Aitkaliyeva, Assel [Materials Science and Engineering Program, Texas A and M University, College Station, Texas 77843 (United States); David Theodore, N. [CHD-Fab, Freescale Semiconductor, Inc., Chandler, Arizona 85224 (United States); Xie Guoqiang [Institute for Materials Research, Tohoku University, Sendai 980-8577 (Japan)

2012-07-23T23:59:59.000Z

407

SOLAR WIND HELIUM ABUNDANCE AS A FUNCTION OF SPEED AND HELIOGRAPHIC LATITUDE: VARIATION THROUGH A SOLAR CYCLE  

E-Print Network (OSTI)

SOLAR WIND HELIUM ABUNDANCE AS A FUNCTION OF SPEED AND HELIOGRAPHIC LATITUDE: VARIATION THROUGH A SOLAR CYCLE Justin C. Kasper,1 Michael L. Stevens, and Alan J. Lazarus Kavli Institute for Astrophysics of the variation of the relative abundance of helium to hydrogen in the solar wind as a function of solar wind

Richardson, John

408

Experimental measurement methods and data on irradiation of functional design materials by helium ions in linear accelerator  

E-Print Network (OSTI)

The experimental research on the irradiation of the functional design materials by the Helium ions in the linear accelerator is conducted. The experimental measurements techniques and data on the irradiation of the functional design materials by the Helium ions with the energy up to 4 MeV, including the detailed scheme of experimental measurements setup, are presented. The new design of accelerating structure of the IH-type such as POS-4, using the method of alternate-phase focusing with the step-by-step change of the synchronous phase along the focusing periods in a linear accelerator, is developed with the aim to irradiate the functional design materials by the Helium ions. The new design of the injector of the charged Helium ions with the energy of 120 KeV at the output of an accelerating tube and the accelerating structure of the type of POS-4 for the one time charged Helium ions acceleration in the linear accelerator are researched and developed. The special chamber for the irradiation of functional design materials by the Helium ions is also created. In the process of experiment, the temperature of a sample, the magnitude of current of Helium ions beam and the irradiation dose of sample are measured precisely. The experimental measurement setup and techniques are fully tested and optimized in the course of the research on the electro-physical properties of irradiated samples and the thermal-desorption of Helium ions in a wide range of temperatures

R. A. Anokhin; V. N. Voyevodin; S. N. Dubnyuk; A. M. Egorov; B. V. Zaitsev; A. F. Kobets; O. P. Ledenyov; K. V. Pavliy; V. V. Ruzhitsky; G. D. Tolstolutskaya

2013-09-03T23:59:59.000Z

409

Nuclear reactor construction with bottom supported reactor vessel  

DOE Patents (OSTI)

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 structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

Sharbaugh, John E. (Bullskin Township, Fayette County, PA)

1987-01-01T23:59:59.000Z

410

HOMOGENEOUS NUCLEAR REACTOR  

SciTech Connect

This homogeneous reactor comprises a core occupied by a solution of a fissile material in a moderator liquid and a breeder region enclosing the core and having a suspension of fertile material in the same moderator liquid. There is communication between the core and breeder to allow mass transfer and pressure equalization between the regions. The zones each have a separate circuit for removing heat by a mixer chamber situated inside the reactor vessel. The effluents coming from the two regions are mixed and led to a common device for separation into a clear solution and suspension, which are each led back to its corresponding circuit. To control the relative concentration of the two regions, an evaporator is provided separating a part of the moderator liquid from the solution occupying the core, the condensed separated moderator liquid being led into the breeder region. (NPO)

1960-07-11T23:59:59.000Z

411

Fusion reactor pumped laser  

DOE Patents (OSTI)

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.

Jassby, D.L.

1987-09-04T23:59:59.000Z

412

AIR COOLED NEUTRONIC REACTOR  

DOE Patents (OSTI)

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.

Fermi, E.; Szilard, L.

1958-05-27T23:59:59.000Z

413

Nuclear reactor safety device  

DOE Patents (OSTI)

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.

Hutter, E.

1983-08-15T23:59:59.000Z

414

Spherical torus fusion reactor  

DOE Patents (OSTI)

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.

Peng, Yueng-Kay M. (Oak Ridge, TN)

1989-01-01T23:59:59.000Z

415

HOMOGENEOUS NUCLEAR REACTOR  

DOE Patents (OSTI)

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.

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

1959-02-17T23:59:59.000Z

416

Water reactor fuel cladding  

Science Conference Proceedings (OSTI)

This patent describes a nuclear reactor fuel element cladding tube. It comprises: an outer cylindrical layer of a first zirconium alloy selected from the group consisting of Zircaloy-2 and Zircaloy-4; an inner cylindrical layer of a second zirconium alloy consisting essentially of about 0.19 to 0.6 wt.% tin, about 0.19 to less than 0.5 wt.% iron, about 100 to 700 ppm oxygen, less than 2000 ppm total impurities, and the remainder essentially zirconium; the inner layer characterized by aqueous corrosion resistance substantially the same as the first zirconium alloy; the inner layer characterized by improved resistance to PCI crack propagation under reactor operating conditions compared to the first zirconium alloy and substantially the same PCI crack propagation resistance compared to unalloyed zirconium; and the inner cylindrical layer is metallurgically bonded to the outer layer.

Foster, J.P.; McDonald, S.G.

1990-06-12T23:59:59.000Z

417

Power Reactor Decommissioning Experience  

Science Conference Proceedings (OSTI)

During the past two decades the NRC regulated nuclear industry has encountered and dealt with a diverse range of political, financial and technological challenges while decommissioning its nuclear facilities. During that time, the decommissioning of nuclear facilities has evolved into a mature industry in the United States with a number of large power reactors successfully decommissioned and their NRC licenses terminated. One of the challenges discussed in this report is site release standards, required ...

2011-07-08T23:59:59.000Z

418

BOILER-SUPERHEATED REACTOR  

DOE Patents (OSTI)

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.)

Heckman, T.P.

1961-05-01T23:59:59.000Z

419

Exhaust gas reactor  

SciTech Connect

A reactor for the oxidation of unburned and partially burned components in the exhaust gas from an internal combustion engine comprising a chamber which is substantially circular in cross sections perpendicular to its axis, one or more inlet pipes which pass a mixture of exhaust gas and air substantially tangentially into the chamber near to one end thereof, and an outlet pipe near to the other end of the chamber and which is so arranged that exhaust gas leaves the chamber substantially tangentially. The tangential inlet and tangential outlet of gases minimizes energy losses in the gas passing through the reactor. The ratio of the cross-sectional areas of the inlet pipe(s) to reactor chamber is preferably from 1:9 to 25:36, and similar ranges of crosssectional area ratios are preferred for the outlet pipe and chamber. The ratio of the length of the reaction chamber to diameter is preferably from 1:1 to 4:1. The chamber may be cylindrical or divergent from inlet end to outlet end and may be thermally insulated.

Camarsa, M.; Cocchiara, F.; Garcea, G.P.

1981-11-24T23:59:59.000Z

420

Modeling Chemical Reactors I: Quiescent Reactors  

E-Print Network (OSTI)

We introduce a fully generalized quiescent chemical reactor system in arbitrary space $\\vdim =1,2$ or 3, with $n\\in\\mathbb{N}$ chemical constituents $\\alpha_{i}$, where the character of the numerical solution is strongly determined by the relative scaling between the local reactivity of species $\\alpha_{i}$ and the local functional diffusivity $\\mathscr{D}_{ij}(\\alpha)$ of the reaction mixture. We develop an operator time-splitting predictor multi-corrector RK--LDG scheme, and utilize $hp$-adaptivity relying only on the entropy $\\mathscr{S}_{\\mathfrak{R}}$ of the reactive system $\\mathfrak{R}$. This condition preserves these bounded nonlinear entropy functionals as a necessarily enforced stability condition on the coupled system. We apply this scheme to a number of application problems in chemical kinetics; including a difficult classical problem arising in nonequilibrium thermodynamics known as the Belousov-Zhabotinskii reaction where we utilize a concentration-dependent diffusivity tensor $\\mathscr{D}_{ij}(\\alpha)$, in addition to solving a simple equilibrium problem in order to evaluate the numerical error behavior.

C. E. Michoski; J. A. Evans; P. G. Schmitz

2010-12-28T23:59:59.000Z

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


421

Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review  

SciTech Connect

In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path.

Lund, A.L.

1997-11-01T23:59:59.000Z

422

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

DOE Green Energy (OSTI)

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 and thus reduce three temperature losses in the system associated with (1) heat transfer from the fuel to the reactor coolant, (2) temperature rise across the reactor core, and (3) heat transfer across the heat exchangers between the reactor and H2 production plant. Lowering the peak reactor temperatures and thus reducing the high-temperature materials requirements may make the AHTR the enabling technology for low-cost nuclear hydrogen production.

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

2004-10-06T23:59:59.000Z

423

Design guide for category V reactors transient reactors  

SciTech Connect

The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category V reactor structures, components, and systems.

Brynda, W J; Karol, R C; Lobner, P R; Powell, R W; Straker, E A

1979-03-01T23:59:59.000Z

424

A Concept for A Dark Matter Detector Using Liquid Helium-4  

E-Print Network (OSTI)

Direct searches for light dark matter particles (mass $<10$ GeV) are especially challenging because of the low energies transferred in elastic scattering to typical heavy nuclear targets. We investigate the possibility of using liquid Helium-4 as a target material, taking advantage of the favorable kinematic matching of the Helium nucleus to light dark matter particles. Monte Carlo simulations are performed to calculate the charge, scintillation, and triplet helium molecule signals produced by recoil He ions, for a variety of energies and electric fields. We show that excellent background rejection can be achieved based on the ratios between different signal channels. We also present some concepts for a liquid-helium-based dark matter detector. Key to the proposed approach is the use of a large electric field to extract electrons from the event site, and the amplification of this charge signal, through proportional scintillation, liquid electroluminescence, or roton emission. The sensitivity of the proposed detector to light dark matter particles is estimated for various electric fields and light collection efficiencies.

Wei Guo; Daniel N. McKinsey

2013-02-03T23:59:59.000Z

425

Electron-impact excitation-autoionization of helium in the S-wave limit  

Science Conference Proceedings (OSTI)

Excitation of the autoionizing states of helium by electron impact is shown in calculations in the s-wave limit to leave a clear signature in the singly differential cross section for the (e,2e) process. It is suggested that such behavior should be seen generally in (e,2e) experiments on atoms that measure the single differential cross section.

Horner, Daniel A.; McCurdy, C. William; Rescigno, Thomas N.

2004-10-01T23:59:59.000Z

426

RECONCILING THE GALACTIC BULGE TURNOFF AGE DISCREPANCY WITH ENHANCED HELIUM ENRICHMENT  

SciTech Connect

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.

Nataf, David M.; Gould, Andrew P. [Department of Astronomy, Ohio State University, 140 West 18th Avenue, Columbus, OH 43210 (United States)

2012-06-01T23:59:59.000Z

427

Helium Adsorption in Silica Aerogel near the Liquid-Vapor Critical Point  

E-Print Network (OSTI)

We have investigated the adsorption and desorption of helium near its liquid-vapor critical point in silica aerogels with porosities between 95 % and 98%. We used a capacitive measurement technique which allowed us to probe the helium density inside the aerogel directly, even though the samples were surrounded by bulk helium. The aerogel’s very low thermal conductivity resulted in long equilibration times so we monitored the pressure and the helium density, both inside the aerogel and in the surrounding bulk, and waited at each point until all had stabilized. Our measurements were made at temperatures far from the critical point, where a well defined liquid-vapor interface exists, and at temperatures up to the bulk critical point. Hysteresis between adsorption and desorption isotherms persisted to temperatures close to the liquid-vapor critical point and there was no sign of an equilibrium liquid-vapor transition once the hysteresis disappeared. Many features of our isotherms can be described in terms of capillary condensation, although this picture becomes less applicable as the liquid-vapor critical point is approached and it is unclear how it can be applied to aerogels, whose tenuous structure includes a wide range of length scales. I.

Tobias Herman; James Day; John Beamish

2008-01-01T23:59:59.000Z

428

A temperature-controlled device for volumetric measurements of Helium adsorption in porous  

E-Print Network (OSTI)

a set-up for studying adsorption of helium in silica aerogels, where the adsorbed amount is easily- dation experiments and a first application to aerogels. This device is well adapted to study hysteresis, such as grafoil, carbon nanotubes, or silica glasses. The case of silica aerogels has attracted attention

Paris-Sud XI, Université de

429

Helium adsorption in silica aerogel near the liquid-vapor critical point  

E-Print Network (OSTI)

We have investigated the adsorption and desorption of helium near its liquid-vapor critical point in silica aerogels with porosities between 95% and 98%. We used a capacitive measurement technique which allowed us to probe the helium density inside the aerogel directly, even though the samples were surrounded by bulk helium. The aerogel's very low thermal conductivity resulted in long equilibration times so we monitored the pressure and the helium density, both inside the aerogel and in the surrounding bulk, and waited at each point until all had stabilized. Our measurements were made at temperatures far from the critical point, where a well defined liquid-vapor interface exists, and at temperatures up to the bulk critical point. Hysteresis between adsorption and desorption isotherms persisted to temperatures close to the liquid-vapor critical point and there was no sign of an equilibrium liquid-vapor transition once the hysteresis disappeared. Many features of our isotherms can be described in terms of capillary condensation, although this picture becomes less applicable as the liquid-vapor critical point is approached and it is unclear how it can be applied to aerogels, whose tenuous structure includes a wide range of length scales.

Tobias Herman; James Day; John Beamish

2005-05-18T23:59:59.000Z

430

Direct evidence of mismatching effect on H emission in laser-induced atmospheric helium gas plasma  

Science Conference Proceedings (OSTI)

A time-resolved orthogonal double pulse laser-induced breakdown spectroscopy (LIBS) with helium surrounding gas is developed for the explicit demonstration of time mismatch between the passage of fast moving impurity hydrogen atoms and the formation of thermal shock wave plasma generated by the relatively slow moving major host atoms of much greater masses ablated from the same sample. Although this so-called 'mismatching effect' has been consistently shown to be responsible for the gas pressure induced intensity diminution of hydrogen emission in a number of LIBS measurements using different ambient gases, its explicit demonstration has yet to be reported. The previously reported helium assisted excitation process has made possible the use of surrounding helium gas in our experimental set-up for showing that the ablated hydrogen atoms indeed move faster than the simultaneously ablated much heavier major host atoms as signaled by the earlier H emission in the helium plasma generated by a separate laser prior to the laser ablation. This conclusion is further substantiated by the observed dominant distribution of H atoms in the forward cone-shaped target plasma.

Zener Sukra Lie; Koo Hendrik Kurniawan [Research Center of Maju Makmur Mandiri Foundation, 40 Srengseng Raya, Kembangan, Jakarta Barat 11630 (Indonesia); May On Tjia [Research Center of Maju Makmur Mandiri Foundation, 40 Srengseng Raya, Kembangan, Jakarta Barat 11630 (Indonesia); Physics of Magnetism and Photonics Group, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, 10 Ganesha, Bandung 40132 (Indonesia); Rinda, Hedwig [Department of Computer Engineering, Bina Nusantara University, 9 K.H. Syahdan, Jakarta 14810 (Indonesia); Suliyanti, Maria Margaretha [Research Center for Physics, Indonesia Institute of Sciences, Kawasan PUSPIPTEK, Serpong, Tangerang Selatan 15314, Banten (Indonesia); Syahrun Nur Abdulmadjid; Nasrullah Idris [Department of Physics, Faculty of Mathematics and Natural Sciences, Syiah Kuala University, Darussalam, Banda Aceh 23111, NAD (Indonesia); Alion Mangasi Marpaung [Department of Physics, Faculty of Mathematics and Natural Sciences, Jakarta State University, Rawamangun, Jakarta 12440 (Indonesia); Marincan Pardede [Department of Electrical Engineering, University of Pelita Harapan, 1100 M.H. Thamrin Boulevard, Lippo Village, Tangerang 15811 (Indonesia); Jobiliong, Eric [Department of Industrial Engineering, University of Pelita Harapan, 1100 M.H. Thamrin Boulevard, Lippo Village, Tangerang 15811 (Indonesia); Muliadi Ramli [Department of Chemistry, Faculty of Mathematics and Natural Sciences, Syiah Kuala University, Darussalam, Banda Aceh 23111, NAD (Indonesia); Heri Suyanto [Department of Physics, Faculty of Mathematics and Natural Sciences, Udayana University, Kampus Bukit Jimbaran, Denpasar 80361, Bali (Indonesia); Fukumoto, Kenichi; Kagawa, Kiichiro [Research Institute of Nuclear Engineering, University of Fukui, Fukui 910-8507 (Japan)

2013-02-07T23:59:59.000Z

431

Heavy Water and Graphite Reactors - Reactors designed/built by...  

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

experiments, necessary to achieve higher precision for the determination of reactor power distribution patterns, effect of non-uniform void distributions, kinetic behavior,...

432

Fusion blankets for catalyzed D--D and D--/sup 3/He reactors  

SciTech Connect

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He/sup 3/ fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) ..beta.. non-circular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He/sup 3/ Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphenyl coolant.

Fillo, J.A.; Powell, J.R.

1977-01-01T23:59:59.000Z

433

Fusion blankets for catalyzed D--D and D--He/sup 3/ reactors  

SciTech Connect

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He/sup 3/ fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) ..beta.. noncircular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He/sup 3/ Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphynyl coolant.

Fillo, J.A.; Powell, J.R.

1977-01-01T23:59:59.000Z

434

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

Science Conference Proceedings (OSTI)

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.

Jonathan A. Webb; Brian Gross

2011-02-01T23:59:59.000Z

435

Preliminary Benchmark Evaluation of Japan’s High Temperature Engineering Test Reactor  

SciTech Connect

A benchmark model of the initial fully-loaded start-up core critical of Japan’s High Temperature Engineering Test Reactor (HTTR) was developed to provide data in support of ongoing validation efforts of the Very High Temperature Reactor Program using publicly available resources. The HTTR is a 30 MWt test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. The benchmark was modeled using MCNP5 with various neutron cross-section libraries. An uncertainty evaluation was performed by perturbing the benchmark model and comparing the resultant eigenvalues. The calculated eigenvalues are approximately 2-3% greater than expected with an uncertainty of ±0.70%. The primary sources of uncertainty are the impurities in the core and reflector graphite. The release of additional HTTR data could effectively reduce the benchmark model uncertainties and bias. Sensitivity of the results to the graphite impurity content might imply that further evaluation of the graphite content could significantly improve calculated results. Proper characterization of graphite for future Next Generation Nuclear Power reactor designs will improve computational modeling capabilities. Current benchmarking activities include evaluation of the annular HTTR cores and assessment of the remaining start-up core physics experiments, including reactivity effects, reactivity coefficient, and reaction-rate distribution measurements. Long term benchmarking goals might include analyses of the hot zero-power critical, rise-to-power tests, and other irradiation, safety, and technical evaluations performed with the HTTR.

John Darrell Bess

2009-05-01T23:59:59.000Z

436

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

Science Conference Proceedings (OSTI)

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)

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

2006-07-01T23:59:59.000Z

437

Nuclear Reactor Severe Accident Experiments  

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

Nuclear Reactor Severe Accident Experiments Nuclear Reactor Severe Accident Experiments Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Nuclear Reactor Severe Accident Experiments 1 2 3 4 5 6 7 We perform experiments simulating reactor core melt phenomena in which molten core debris ("corium") erodes the concrete floor of a containment building. This occurred during the Fukushima nuclear power plant accident though the extent of concrete damage is yet unknown. This video shows the top view of a churning molten pool of uranium oxide at 2000°C (3600°F) seen during an experiment at Argonne. Corium behaves much like lava.

438

Designing Reactors to Facilitate Decommissioning  

SciTech Connect

Critics of nuclear power often cite issues with tail-end-of-the-fuel-cycle activities as reasons to oppose the building of new reactors. In fact, waste disposal and the decommissioning of large nuclear reactors have proven more challenging than anticipated. In the early days of the nuclear power industry the design and operation of various reactor systems was given a great deal of attention. Little effort, however, was expended on end-of-the-cycle activities, such as decommissioning and disposal of wastes. As early power and test reactors have been decommissioned difficulties with end-of-the-fuel-cycle activities have become evident. Even the small test reactors common at the INEEL were not designed to facilitate their eventual decontamination, decommissioning, and dismantlement. The results are that decommissioning of these facilities is expensive, time consuming, relatively hazardous, and generates large volumes of waste. This situation clearly supports critics concerns about building a new generation of power reactors.

Richard H. Meservey

2006-06-01T23:59:59.000Z

439

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

Science Conference Proceedings (OSTI)

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

Fauzia, A. F.; Waris, A.; Novitrian [Bosscha Laboratory, Department of Physics, Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, INDONESIA Jl. Ganesa 10 Bandung 40132 (Indonesia)

2010-06-22T23:59:59.000Z

440

Acceptability of reactors in space  

SciTech Connect

Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it does not appear that reactors add measurably to the risk associated with the Space Transportation System.

Buden, D.

1981-01-01T23:59:59.000Z

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441

Acceptability of reactors in space  

SciTech Connect

Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System.

Buden, D.

1981-04-01T23:59:59.000Z

442

Survey of helium in soils and soil gases and mercury in soils at Roosevelt Hot Springs Known Geothermal Resource Area, Utah  

SciTech Connect

The concentrations of helium and mercury in soils and of helium in soil gases were surveyed in part of the Roosevelt Hot Springs Known Geothermal Resource Area to see what relationship helium and mercury concentrations might have to geothermal features of the area. High concentrations of helium occurred over the producing geothermal field, in an area of high temperature gradients. Low concentrations of helium in soils occurred over an area of visible hydrotheormal activity. High concentrations of mercury coincided with areas of high thermal gradients and low resistivity.

Hinkle, M.E.

1980-01-01T23:59:59.000Z

443

Transport Reactor Facility  

SciTech Connect

The Morgantown Energy Technology Center (METC) is currently evaluating hot gas desulfurization (HGD)in its on-site transport reactor facility (TRF). This facility was originally constructed in the early 1980s to explore advanced gasification processes with an entrained reactor, and has recently been modified to incorporate a transport riser reactor. The TRF supports Integrated Gasification Combined Cycle (IGCC) power systems, one of METC`s advanced power generation systems. The HGD subsystem is a key developmental item in reducing the cost and increasing the efficiency of the IGCC concept. The TRF is a unique facility with high-temperature, high-pressure, and multiple reactant gas composition capability. The TRF can be configured for reacting a single flow pass of gas and solids using a variety of gases. The gas input system allows six different gas inputs to be mixed and heated before entering the reaction zones. Current configurations allow the use of air, carbon dioxide, carbon monoxide, hydrogen, hydrogen sulfide, methane, nitrogen, oxygen, steam, or any mixture of these gases. Construction plans include the addition of a coal gas input line. This line will bring hot coal gas from the existing Fluidized-Bed Gasifier (FBG) via the Modular Gas Cleanup Rig (MGCR) after filtering out particulates with ceramic candle filters. Solids can be fed either by a rotary pocket feeder or a screw feeder. Particle sizes may range from 70 to 150 micrometers. Both feeders have a hopper that can hold enough solid for fairly lengthy tests at the higher feed rates, thus eliminating the need for lockhopper transfers during operation.

Berry, D.A.; Shoemaker, S.A.

1996-12-31T23:59:59.000Z

444

NEUTRONIC REACTOR CHARGING AND DISCHARGING  

DOE Patents (OSTI)

A method and arrangement is presented for removing a fuel element from a neutronic reactor tube through which a liquid coolant is being circulaled. The fuel element is moved into a section of the tube beyond the reactor proper, and then the coolant in the tube between the fuel element and the reactor proper is frozen, so that the fuel element may be removed from the tube without loss of the coolant therein. The method is particularly useful in the case of a liquid metal- cooled reactor.

Zinn, W.H.

1959-07-14T23:59:59.000Z

445

Reactor & Nuclear Systems Publications | ORNL  

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

Publications and Reports NSED Monthly Reports Reactor and Nuclear Systems Publications 2013 Publications 2012 Publications 2011 Publications 2010 and Older Publications Nuclear...

446

Mirror Advanced Reactor Study (MARS)  

DOE Green Energy (OSTI)

Progress in a two year study of a 1200 MWe commercial tandem mirror reactor (MARS - Mirror Advanced Reactor Study) has reached the point where major reactor system technologies are identified. New design features of the magnets, blankets, plug heating systems and direct converter are described. With the innovation of radial drift pumping to maintain low plug density, reactor recirculating power fraction is reduced to 20%. Dominance of radial ion and impurity losses into the halo permits gridless, circular direct converters to be dramatically reduced in size. Comparisons of MARS with the Starfire tokamak design are made.

Logan, B.G.

1983-03-28T23:59:59.000Z

447