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1

Reactor operation safety information document  

SciTech Connect

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

2

Operational Analysis of Multiregional Nuclear Reactor Kinetics  

Science Journals Connector (OSTI)

......Operational Analysis of Multiregional Nuclear Reactor Kinetics NASSAR H. S. HAIDAR...analytically for a multiregional nuclear reactor whose subregions are of arbitrary...Operational Analysis of Multiregional Nuclear Reactor Kinetics NASSAU H. S. HAIDAR......

NASSAR H. S. HAIDAR

1983-05-01T23:59:59.000Z

3

Assessment of typical BWR (boiling water reactor) vessel configurations and examination coverage  

SciTech Connect

Even though boiling water reactors (BWRs) are not susceptible to the kind of incident known as pressurized thermal shock that must be considered in the design and operation of pressurized water reactors, BWR reactor pressure vessels (RPVs) have experienced higher than expected embrittlement caused by fast neutron irradiation. This has required the vessel to be at a higher temperature than originally projected before the plant can be taken to power operation. In addition, many BWR plants have received exemption from the 10-year volumetric nondestructive evaluation (NDE) of the vessel as required by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B PV) Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components,'' because NDE access is severely restricted. Since many RPV welds have not been examined after being placed in service and the potential for service-induced flaws exists, regulatory authorities are looking closely at examination relief requests. BWR reactor vessel examination coverage is typically limited by plant design. Most BWR plants were designed when inservice examination codes were in the early stages of development, and very little consideration was give to designing for NDE access. Consequently, there is restricted access for many areas of the RPV. Since an increase in examination requirements has been placed in ASME B PV Code Section XI in these areas, efforts have begun on a thorough analysis of the vessel weld volumes examined during inservice examination and an evaluation of possibility expanding the RPV examination coverage. Because of these concerns, an investigation of the accessibility of the reactor vessel for NDE was performed to define the present status and to determine the improvements in coverage that can be accomplished in the near future. 7 refs., 9 figs., 4 tabs.

Walker, S.M. (EPRI Nondestructive Evaluation Center, Charlotte, NC (USA)); Feige, E.J.; Ingamells, J.R. (Southwest Research Inst., San Antonio, TX (USA)); Calhoun, G.L.; Davis, J.; Kapoor, A. (Westinghouse Electric Corp., Pittsburgh, PA (USA))

1990-10-01T23:59:59.000Z

4

Digital computer operation of a nuclear reactor  

DOE Patents (OSTI)

A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

Colley, Robert W. (Richland, WA)

1984-01-01T23:59:59.000Z

5

Digital computer operation of a nuclear reactor  

DOE Patents (OSTI)

A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

Colley, R.W.

1982-06-29T23:59:59.000Z

6

Nuclear reactor characteristics and operational history  

Gasoline and Diesel Fuel Update (EIA)

Nuclear > U.S. reactor operation status tables Nuclear > U.S. reactor operation status tables Nuclear Reactor Operational Status Tables Release date: November 22, 2011 Next release date: November 2012 See also: Table 1. Capacity and Generation, Table 2. Ownership Data Table 3. Nuclear Reactor Characteristics and Operational History PDF XLS Plant Name Generator ID Type Reactor Supplier and Model Construction Start Grid Connection Original Expiration Date License Renewal Application License Renewal Issued Extended Expiration Arkansas Nuclear One 1 PWR Babcock&Wilcox, Lower Loop 10/1/1968 8/17/1974 5/20/2014 2/1/2000 6/20/2001 5/20/2034 Arkansas Nuclear One 2 PWR Combustion Eng. 7/1/1971 12/26/1978 7/17/2018 10/15/2003 6/30/2005 7/17/2038

7

Nuclear reactor characteristics and operational history  

U.S. Energy Information Administration (EIA) Indexed Site

Nuclear > U.S. reactor operation status tables Nuclear > U.S. reactor operation status tables Nuclear Reactor Operational Status Tables Release date: November 22, 2011 Next release date: November 2012 See also: Table 2. Ownership Data, Table 3. Characteristics and Operational History Table 1. Nuclear Reactor, State, Type, Net Capacity, Generation, and Capacity Factor PDF XLS Plant/Reactor Name Generator ID State Type 2009 Summer Capacity Net MW(e)1 2010 Annual Generation Net MWh2 Capacity Factor Percent3 Arkansas Nuclear One 1 AR PWR 842 6,607,090 90 Arkansas Nuclear One 2 AR PWR 993 8,415,588 97 Beaver Valley 1 PA PWR 892 7,119,413 91 Beaver Valley 2 PA PWR 885 7,874,151 102 Braidwood Generation Station 1 IL PWR 1,178 9,196,689 89

8

Operational control of boiling water reactor stability  

SciTech Connect

Boiling water reactor cores are susceptible to instabilities, which generate power oscillations. Specific reactor operating practices can provide a mechanism for control of the instability phenomenon. An axial separation of the core into a single-phase region and a two-phase region resolves the influence of axial flux shapes on core stability. This separation provides the means to derive a core stability control that ensures significant reactor stability margin. The control is achieved by maintaining the core average bulk coolant saturation elevation above a predetermined axial plane. The control can be reliably and efficiently implemented during reactor operations. Analysis demonstrates that variations in parameters important to stability have only secondary influences on stability margin when the control is in effect. Actual plant experience with a large commercial boiling water reactor confirms the capabilities of this stability control in an operational setting.

Mowry, C.M. [PECO Energy, Wayne, PA (United States); Nir, I. [Entergy Operations, Jackson, MS (United States); Newkirk, D.W. [GE Nuclear Energy, San Jose, CA (United States)

1995-03-01T23:59:59.000Z

9

Nuclear reactor characteristics and operational history  

Gasoline and Diesel Fuel Update (EIA)

1. Capacity and Generation, Table 3. Characteristics and Operational History 1. Capacity and Generation, Table 3. Characteristics and Operational History Table 2. U.S. Nuclear Reactor Ownership Data PDF XLS Plant/Reactor Name Generator ID Utility Name - Operator Owner Name % Owned Arkansas Nuclear One 1 Entergy Arkansas Inc Entergy Arkansas Inc 100 Arkansas Nuclear One 2 Entergy Arkansas Inc Entergy Arkansas Inc 100 Beaver Valley 1 FirstEnergy Nuclear Operating Company FirstEnergy Nuclear Generation Corp 100 Beaver Valley 2 FirstEnergy Nuclear Operating Company FirstEnergy Nuclear Generation Corp 100 Braidwood Generation Station 1 Exelon Nuclear Exelon Nuclear 100 Braidwood Generation Station 2 Exelon Nuclear Exelon Nuclear 100 Browns Ferry 1 Tennessee Valley Authority Tennessee Valley Authority 100

10

Table 3. Nuclear Reactor Characteristics and Operational History  

U.S. Energy Information Administration (EIA) Indexed Site

3. Nuclear Reactor Characteristics and Operational History" "Plant Name","Generator ID","Type","Reactor Supplier and Model","Construction Start","Grid Connection","Commercial...

11

CRAD, Nuclear Reactor Facility Operations - December 4, 2014...  

Energy Savers (EERE)

Nuclear Reactor Facility Operations - December 4, 2014 (EA CRAD 31-08, Rev. 0) CRAD, Nuclear Reactor Facility Operations - December 4, 2014 (EA CRAD 31-08, Rev. 0) December 4, 2014...

12

Using reactor operating experience to improve the design of a new Broad Application Test Reactor  

SciTech Connect

Increasing regulatory demands and effects of plant aging are limiting the operation of existing test reactors. Additionally, these reactors have limited capacities and capabilities for supporting future testing missions. A multidisciplinary team of experts developed sets of preliminary safety requirements, facility user needs, and reactor design concepts for a new Broad Application Test Reactor (BATR). Anticipated missions for the new reactor include fuels and materials irradiation testing, isotope production, space testing, medical research, fusion testing, intense positron research, and transmutation doping. The early BATR design decisions have benefited from operating experiences with existing reactors. This paper discusses these experiences and highlights their significance for the design of a new BATR.

Fletcher, C.D.; Ryskamp, J.M.; Drexler, R.L.; Leyse, C.F.

1993-07-01T23:59:59.000Z

13

A GASFLOW analysis of a steam explosion accident in a typical light-water reactor confinement building  

SciTech Connect

Steam over-pressurization resulting from ex-vessel steam explosion (fuel-coolant interaction) may pose a serious challenge to the integrity of a typical light-water reactor confinement building. If the steam generation rate exceeds the removal capacity of the Airborne Activity Confinement System, confinement over pressurization occurs. Thus, there is a large potential for an uncontrolled and unfiltered release of fission products from the confinement atmosphere to the environment at the time of the steam explosion. The GASFLOW computer code was used to analyze the effects of a hypothetical steam explosion and the transport of steam and hydrogen throughout a typical light-water reactor confinement building. The effects of rapid pressurization and the resulting forces on the internal structures and the heat exchanger service bay hatch covers were calculated. Pressurization of the ventilation system and the potential damage to the ventilation fans and high-efficiency particulate air filters were assessed. Because of buoyancy forces and the calculated confinement velocity field, the hydrogen diffuses and mixes in the confinement atmosphere but tends to be transported to its upper region.

Travis, J.R. [ESSI Inc. (United States); Wilson, T.L.; Spore, J.W.; Lam, K.L. [Los Alamos National Lab., NM (United States); Rao, D.V. [SEA Inc. (United States)

1994-09-01T23:59:59.000Z

14

Hydrogasification reactor and method of operating same  

SciTech Connect

The present invention provides a system and method for evaluating effects of process parameters on hydrogasification processes. The system includes a hydrogasification reactor, a pressurized feed system, a hopper system, a hydrogen gas source, and a carrier gas source. Pressurized carbonaceous material, such as coal, is fed to the reactor using the carrier gas and reacted with hydrogen to produce natural gas.

Hobbs, Raymond; Karner, Donald; Sun, Xiaolei; Boyle, John; Noguchi, Fuyuki

2013-09-10T23:59:59.000Z

15

Influence of seeded impurities on the fusion reactor operation  

Science Journals Connector (OSTI)

In order to study the influence of seeded impurities on ITER like reactor operation the COREDIV code has been extended to include the transport of several sputtered and/or injected impurities. In the COREDIV c...

R. Stankiewicz; R. Zagórski

2004-03-01T23:59:59.000Z

16

Simulator platform for fast reactor operation and safety technology demonstration  

SciTech Connect

A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)

2012-07-30T23:59:59.000Z

17

On the operator action analysis to reduce operational risk in research reactors  

Science Journals Connector (OSTI)

Abstract Human errors during operation and the resulting increase in operational risk are major concerns for nuclear reactors, just as they are for all industries. Additionally, human reliability analysis together with probabilistic risk analysis is a key element in reducing operational risk. The purpose of this paper is to analyze human reliability using appropriate methods for the probabilistic representation and calculation of human error to be used alongside probabilistic risk analysis in order to reduce the operational risk of the reactor operation. We present a technique for human error rate prediction and standardized plant analysis risk. Human reliability methods have been utilized to quantify different categories of human errors, which have been applied extensively to nuclear power plants. The Tehran research reactor is selected here as a case study, and after consultation with reactor operators and engineers human errors have been identified and adequate performance shaping factors assigned in order to calculate accurate probabilities of human failure.

Ramin Barati; Saeed Setayeshi

2014-01-01T23:59:59.000Z

18

Operation of staged membrane oxidation reactor systems  

SciTech Connect

A method of operating a multi-stage ion transport membrane oxidation system. The method comprises providing a multi-stage ion transport membrane oxidation system with at least a first membrane oxidation stage and a second membrane oxidation stage, operating the ion transport membrane oxidation system at operating conditions including a characteristic temperature of the first membrane oxidation stage and a characteristic temperature of the second membrane oxidation stage; and controlling the production capacity and/or the product quality by changing the characteristic temperature of the first membrane oxidation stage and/or changing the characteristic temperature of the second membrane oxidation stage.

Repasky, John Michael

2012-10-16T23:59:59.000Z

19

Pressurized reactor system and a method of operating the same  

DOE Patents (OSTI)

A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Superatmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gassification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor.

Isaksson, Juhani M. (Karhula, FI)

1996-01-01T23:59:59.000Z

20

Pressurized reactor system and a method of operating the same  

DOE Patents (OSTI)

A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Super-atmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gasification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor. 2 figs.

Isaksson, J.M.

1996-06-18T23:59:59.000Z

Note: This page contains sample records for the topic "reactor operators typically" 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

A Controllability Study of TRUMOX Fuel for Load Following Operation in a CANDU-900 Reactor.  

E-Print Network (OSTI)

?? The CANDU-900 reactor design is an improvement on the current CANDU-6 reactor in the areas of economics, safety of operation and fuel cycle flexibility.… (more)

Trudell, David A

2012-01-01T23:59:59.000Z

22

An evaluation of the ecological consequences of partial-power operation of the K Reactor, SRS  

SciTech Connect

The K Reactor at the Savannah River Site (SRS) shut-down in spring 1988 for maintenance and safety upgrades. Since that time the receiving stream for thermal effluent, Indian Grave Branch and Pen Branch, have undergone a pattern of post-thermal recovery that is typical of other SRS streams following removal of thermal stress. Divesity of fish and aquatic macroinvertebrate communities has increased and available habitats have been colonized by numerous species of herbaceous and woody plants. K Reactor is scheduled to resume operation in 1991 and operate through 1992 without a cooling tower to cool the discharge. It is likely that the reactor will operate at approximately one-third to one-half of full power (800--1200 MW thermal) during this period and effluent temperatures will be substantially lower than earlier operation at full power. Monthly average discharge temperatures at half-power operation will range from approximately 42{degrees}C in winter to 49{degrees}C in summer. The volume of water discharged will not be affected by altered power levels and will average approximately 10--11 m{sup 3}/s. The ecological consequences of this mode of operation on the Indian Grave/Pen Branch stream system have been evaluated.

Gladden, J.B.; Mackey, H.E.; Paller, M.H.; Specht, W.L.; Wike, L.D.; Wilde, E.W.

1991-06-01T23:59:59.000Z

23

L-Reactor Operation Savannah River Plant Aiken, SC  

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

51371 (F.R.) 51371 (F.R.) NOTICES DEPARTMENT OF ENERGY L-Reactor Operation, Savannah River Plant Aiken, South Carolina; Finding of No Significant Impact Monday, August 23, 1982 *36691 The Department of Energy (DOE) proposes to resume operation of L- Reactor at its Savannah River Plant at Aiken, South Carolina, as soon as it is ready for operation, scheduled for October 1983. The environmental impacts of the resumption of operation have been evaluated in an environmental assessment (DOE/EA-0195), prepared in accordance with the National Environmental Policy Act of 1969 (NEPA) as implemented by regulations promulgated by the Council on Environmental Quality (CEQ) (40 CFR Parts 1500 -1508, November 1978) and DOE implementing guidelines (45 FR 20694, March 28, 1980). Based on the analysis in the assessment, DOE has

24

13 - Generation IV reactor designs, operation and fuel cycle  

Science Journals Connector (OSTI)

Abstract: This chapter looks at Generation IV nuclear reactors, such as the very high-temperature reactor (VHTR), the supercritical water reactor (SCWR), the molten salt reactor (MSR), the sodium-cooled fast reactor (SFR), the lead-cooled fast reactor (LFR) and the gas-cooled fast reactor (GFR). Reactor designs and fuel cycles are also described.

N. Cerullo; G. Lomonaco

2012-01-01T23:59:59.000Z

25

Meeting the reactor operator's information needs using functional analysis  

SciTech Connect

Since the accident at Three Mile Island, many ideas have been proposed for assisting the reactor operator during emergency situations. However, some of the suggested remedies do not alleviate an important shortcoming of the TMI control room: the operators were not presented with the information they needed in a manner which would allow prompt diagnosis of the problem. To address this problem, functional analysis is being applied at the LOFT facility to ensure that the operator's information needs are being met in his procedures and graphic displays. This paper summarizes the current applications of functional analysis at LOFT.

Nelson, W.R.; Clark, M.T.

1980-01-01T23:59:59.000Z

26

Effects of an Advanced Reactor’s Design, Use of Automation, and Mission on Human Operators  

SciTech Connect

The roles, functions, and tasks of the human operator in existing light water nuclear power plants (NPPs) are based on sound nuclear and human factors engineering (HFE) principles, are well defined by the plant’s conduct of operations, and have been validated by years of operating experience. However, advanced NPPs whose engineering designs differ from existing light-water reactors (LWRs) will impose changes on the roles, functions, and tasks of the human operators. The plans to increase the use of automation, reduce staffing levels, and add to the mission of these advanced NPPs will also affect the operator’s roles, functions, and tasks. We assert that these factors, which do not appear to have received a lot of attention by the design engineers of advanced NPPs relative to the attention given to conceptual design of these reactors, can have significant risk implications for the operators and overall plant safety if not mitigated appropriately. This paper presents a high-level analysis of a specific advanced NPP and how its engineered design, its plan to use greater levels of automation, and its expanded mission have risk significant implications on operator performance and overall plant safety.

Jeffrey C. Joe; Johanna H. Oxstrand

2014-06-01T23:59:59.000Z

27

Non-Power Reactor Operator Licensing Examiner Standards. Revision 1  

SciTech Connect

The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience.

NONE

1995-06-01T23:59:59.000Z

28

Modeling of fixed bed methanation reactor for syngas production: Operating window and performance characteristics  

Science Journals Connector (OSTI)

Abstract The present work focuses on the development of phenomenological model for the bio-syngas to methane conversion process. One dimensional heterogeneous and pseudo-homogeneous model were simulated for a typical pilot plant scale fixed bed methanator processing 55 mol/h of CO (total molar flow rate of 310 mol/h) with inlet composition of H2/CO = 3, CO2/CO = 1, CH4/CO = 0.5 at 550 K and 1 atm. Performance of the fixed bed reactor at different operating conditions like CO2/CO ratio, H2/CO ratio, effect of H2O in the feed was studied. It was found that for feeds that were not pre-enriched with hydrogen, presence of water and water gas shift activity was found to decrease the catalyst inventory substantially. CO2 in the inlet feed stream would help to decrease the temperature due to dilution effect and more importantly, can be chosen to maximize methane yield per mole of CO converted. Further, the model was simulated to predict the performance characteristics of reactor with a mixture containing two types of catalyst, one of them being specifically added to increase H2/CO ratio in feed through water gas shift reaction. The work also laid the importance of incorporating pore diffusion and external mass transfer locally in the computation of actual catalyst inventory and reactor volume. The work was useful in selection of operating window and assessing the various viable options for an industrial reactor. The model developed will serve in selection of operability window for commercialization of substitute natural gas synthesis (SNG) process.

Naren Rajan Parlikkad; Stéphane Chambrey; Pascal Fongarland; Nouria Fatah; Andrei Khodakov; Sandra Capela; Olivier Guerrini

2013-01-01T23:59:59.000Z

29

Changes in the mechanical properties of Hastelloy X when exposed to a typical gas-cooled reactor environment  

SciTech Connect

The helium used in a gas-cooled reactor will contain small amounts of H/sub 2/, CO, CH/sub 4/, H/sub 2/O, and N/sub 2/ which can lead to oxidation and carburization/decarburization of structural materials. Long-term creep tests were run on Hastelloy X to 30,000 h at 649 to 871/sup 0/C. It was found that extensive carburization occurred, the minimum creep rate and time to rupture were equal in air and impure helium environments, and the fracture strain was less in helium than in air. Thermal exposure in the temperature range of 538 to 871/sup 0/C resulted in the reduction of ductility in impact and tensile tests at ambient temperature, and this reduction was greater when the exposure was in impure helium rather than in air. A modified alloy with lower chromium and 2% titanium resisted carburization.

McCoy, H.E. Jr.

1981-01-01T23:59:59.000Z

30

Operation of CANDU power reactor in thorium self-sufficient fuel cycle  

Science Journals Connector (OSTI)

This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations...233U and mode of operation in self-sufficient cycle. For the mode of accumulation of 233U it was a...

B. R. Bergelson; A. S. Gerasimov; G. V. Tikhomirov

2007-02-01T23:59:59.000Z

31

Method for loading, operating, and unloading a ball-bed nuclear reactor  

SciTech Connect

This patent describes a method of operating a ball-bed nuclear reactor with fuel element balls. Some have a fissionable material content different from that of others of the balls. It consists of: initially partly filling a reactor core with fuel balls of sufficient fissionable material content for establishing criticality and a desired level of power production at the completion of the partial filling and then, without any further filling of the reactor cavern, starting reactor operation; thereafter without any removal of fuel balls from the reactor cavern, filling fuel balls continually or in groups at relatively short intervals into the reactor cavern during increasing burning up of the fuel balls already, for compensation of the diminishing fissionable material content of the reactor core constituted by the fuel balls until a final total quantity of filling is reached; after the final filling quantity is reached and burning up has occurred, shutting down the reactor, cooling it off, releasing the pressure in the cavern, and thereafter unloading all the fuel balls from the reactor cavern, unloading being begun when the reactor is shut down and being completed before the reactor is restarted.

Teuchert, E.; Haas, K.A.; Gerwin, H.

1987-09-22T23:59:59.000Z

32

A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts  

SciTech Connect

This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

Jacques V Hugo; David I Gertman; Jeffrey C Joe

2014-08-01T23:59:59.000Z

33

Pressurized fluidized bed reactor and a method of operating the same  

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.

1996-02-20T23:59:59.000Z

34

Biological processing in oscillatory baffled reactors: operation, advantages and potential  

Science Journals Connector (OSTI)

...2 Chemical Engineering and Advanced...pharmaceuticals, fuel, health products...scale-up| 1. Introduction Bioprocessing...e.g. biodiesel formation...displace fossil fuels. J. Ind...base-catalysed biodiesel production...reactors. Fuel Processing...Chemical reaction engineering, 3rd edn...

2013-01-01T23:59:59.000Z

35

Work Domain Analysis of a Predecessor Sodium-cooled Reactor as Baseline for AdvSMR Operational Concepts  

SciTech Connect

This report presents the results of the Work Domain Analysis for the Experimental Breeder Reactor (EBR-II). This is part of the phase of the research designed to incorporate Cognitive Work Analysis in the development of a framework for the formalization of an Operational Concept (OpsCon) for Advanced Small Modular Reactors (AdvSMRs). For a new AdvSMR design, information obtained through Cognitive Work Analysis, combined with human performance criteria, can and should be used in during the operational phase of a plant to assess the crew performance aspects associated with identified AdvSMR operational concepts. The main objective of this phase was to develop an analytical and descriptive framework that will help systems and human factors engineers to understand the design and operational requirements of the emerging generation of small, advanced, multi-modular reactors. Using EBR-II as a predecessor to emerging sodium-cooled reactor designs required the application of a method suitable to the structured and systematic analysis of the plant to assist in identifying key features of the work associated with it and to clarify the operational and other constraints. The analysis included the identification and description of operating scenarios that were considered characteristic of this type of nuclear power plant. This is an invaluable aspect of Operational Concept development since it typically reveals aspects of future plant configurations that will have an impact on operations. These include, for example, the effect of core design, different coolants, reactor-to-power conversion unit ratios, modular plant layout, modular versus central control rooms, plant siting, and many more. Multi-modular plants in particular are expected to have a significant impact on overall OpsCon in general, and human performance in particular. To support unconventional modes of operation, the modern control room of a multi-module plant would typically require advanced HSIs that would provide sophisticated operational information visualization, coupled with adaptive automation schemes and operator support systems to reduce complexity. These all have to be mapped at some point to human performance requirements. The EBR-II results will be used as a baseline that will be extrapolated in the extended Cognitive Work Analysis phase to the analysis of a selected advanced sodium-cooled SMR design as a way to establish non-conventional operational concepts. The Work Domain Analysis results achieved during this phase have not only established an organizing and analytical framework for describing existing sociotechnical systems, but have also indicated that the method is particularly suited to the analysis of prospective and immature designs. The results of the EBR-II Work Domain Analysis have indicated that the methodology is scientifically sound and generalizable to any operating environment.

Ronald Farris; David Gertman; Jacques Hugo

2014-03-01T23:59:59.000Z

36

TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report  

SciTech Connect

Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m/sup 2/ and a surface heat flux of 1 MW/m/sup 2/. The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO/sub 2/ rods. The helium coolant pressure is 5 MPa, entering the module at 297/sup 0/C and exiting at 550/sup 0/C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter.

Conn, R.W.; Ghoniem, N.M.; Firestone, M.A. (eds.)

1986-09-01T23:59:59.000Z

37

Multi-unit Operations in Non-Nuclear Systems: Lessons Learned for Small Modular Reactors  

SciTech Connect

The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. Small modular reactors (SMRs) are one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants and may be operated quite differently. One difference is that multiple units may be operated by a single crew (or a single operator) from one control room. The U.S. Nuclear Regulatory Commission (NRC) is examining the human factors engineering (HFE) aspects of SMRs to support licensing reviews. While we reviewed information on SMR designs to obtain information, the designs are not completed and all of the design and operational information is not yet available. Nor is there information on multi-unit operations as envisioned for SMRs available in operating experience. Thus, to gain a better understanding of multi-unit operations we sought the lesson learned from non-nuclear systems that have experience in multi-unit operations, specifically refineries, unmanned aerial vehicles and tele-intensive care units. In this paper we report the lessons learned from these systems and the implications for SMRs.

OHara J. M.; Higgins, J.; DAgostino, A.

2012-01-17T23:59:59.000Z

38

CRAD, Conduct of Operations- Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR  

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February, 2007 assessment of the Conduct of Operations Program portion of an Operational Readiness Review of the Oak Ridge National Laboratory, High Flux Isotope Reactor.

39

The safe, economical operation of a slightly subcritical reactor and transmutor with a small proton accelerator  

SciTech Connect

This report describes methods in which an accelerator can be used to increase the safety and neutron economy of a power reactor and transmutor of long-lived radioactive wastes, such as minor actinides and fission products, by providing neutrons for its subcritical operation. Instead of the rather large subcriticality of k=0.9--0.95 which we originally proposed for such a transmutor, we propose to use a slightly subcritical reactor, such as k=0.99, which will avoid many of the technical difficulties that are associated with large subcriticality, such as localized power peaking, radiation damage due to the injection of medium-energy protons, the high current accelerator, and the requirement for a long beam-expansion section. We analyzed the power drop that occurred in Phoenix reactor, and show that the operating this reactor in subcritical condition improves its safety.

Takahashi, Hiroshi

1994-04-01T23:59:59.000Z

40

An interpretation of information gained from residence time distribution studies for operation of biological reactors  

E-Print Network (OSTI)

comparison of the same two models, that there was no definitive parameter by which to choose; the single exception being a direct comparison of the predicted to the actual conversions from operating reactors. Comparison itself is a formidable and tinre.... (May 1971) Marlow Lee Dodge, B. A. , Rockford College Directed by: Dr. Robert L. Irvine Most rational designs of biological reactors include the use of mass balances and an assumption of a particular hydraulic descrip- tion such as plug...

Dodge, Marlow Lee

2012-06-07T23:59:59.000Z

Note: This page contains sample records for the topic "reactor operators typically" 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

EIS-0108: L-Reactor Operation, Savannah River Plant, Aiken, South Carolina  

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

This Environmental Impact Statement (EIS) was prepared to provide environmental input into the proposed decision to restart L-Reactor operation at the Savannah River Plant (SRP). The Savannah River Plant is a major U.S. Department of Energy (DOE) installation for the production of defense nuclear materials. The proposed restart of L–Reactor would provide defense nuclear materials (i.e. , plutonium) to wet current and near-term needs for national defense purposes.

42

Experiment operations plan for the MT-4 experiment in the NRU reactor. [PWR  

SciTech Connect

A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the formal operations plan for MT-4 - the fourth materials deformation experiment conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. A major objective of MT-4 was to simulate a pressurized water reactor LOCA that could induce fuel rod cladding deformation and rupture due to a short-term adiabatic transient and a peak fuel cladding temperature of 1200K (1700/sup 0/F).

Russcher, G.E.; Wilson, C.L.; Parchen, L.J.; Marshall, R.K.; Hesson, G.M.; Webb, B.J.; Freshley, M.D.

1983-06-01T23:59:59.000Z

43

The Operation of the Tokamak Fusion Test Reactor Tritium Facility  

Science Journals Connector (OSTI)

Design, Operation, and Maintenance of Tritium System / Proceedings of the Fifth Topical Meeting on Tritium Technology In Fission, Fusion, and Isotopic Applications Belgirate, Italy May 28-June 3, 1995

Charles A. Gentile; James L. Anderson; Paul H. LaMarche

44

Method for the operation of internal combustion engines. [gasification reactor for reforming gasoline  

SciTech Connect

This is a method for the operation of internal combustion engines which is designed to decontaminate the exhaust gases. The method includes: feeding a gasification air stream into a gasification reactor; feeding fuel into the same gasification reactor; combining the fuel with the gasification air into a homogeneous fuel-air mixture in the gasification reactor; and converting the fuel-air mixture by partial combustion into a soot -free reformed gas. Then, the reformed gas is fed from the gasification reactor to a mixer where the reformed gas is mixed with combustion air and the reformed gas-air mixture is fed to the internal combustion engine for further combustion with the result that there is intensive decontamination of the exhaust gases which thereby reduces air pollution. The reformed gas temperature is adjusted low for maximum engine output, and is adjusted higher for lower engine temperatures in order to obtain a reformed gas which is richer in hydrogen and thereby produce exhaust gases which are lower in harmful substances. In reference to the exhaust gases in an internal combustion engine, this method achieves the highest possible degree of decontamination, not only of the carbon monoxide and hydrocarbons , but also of the nitrous oxides in the exhaust gases. Using this method, the internal combustion engine can be operated not only with high-test, no-knock gasoline, but also with cheap, lead-free low octane, straight-run gasoline which is low in aromatics and olefins, which normally do not have no-knock properties, and the internal combustion engine can be operated with the lowest possible fuel consumption. The gasification reactor operates through chemical reaction in the presence of a catalyst. Optionally, this method may include a return of part of the reformed gas to the input of the gasification reactor.

Muhlberg, E.

1980-01-29T23:59:59.000Z

45

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

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

A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February, 2007 assessment of the Conduct of Operations Program in preparation for restart of the Oak Ridge National Laboratory, High Flux Isotope Reactor.

46

(Instrumentation and controls technology and reactor operational safety)  

SciTech Connect

While on vacation, the traveler participated as a co-chairman of a panel of instrumentation and controls specialists visiting nuclear establishments in Europe. The purpose of the visit was to assess the status of instrumentation and controls technology for nuclear power in Europe. A list of the sites visited and the personnel contacted is included in this trip report. The visit was sponsored by Loyola College working under contract to the National Science Foundation. All costs were paid by Loyola College, for whom the traveler was a consultant. This was an outside activity approved by DOE. The traveler was surprised by the high level of automaton present in the German Konvoi nuclear power plants built by Siemens AG KWU. The claim was that this was done to improve the safety of the plant by keeping the operator out of the loop'' for the first 30 minutes of some transients or accidents. The traveler was also surprised by the high level of man-machine interface R D in the USSR.

White, J.D.

1990-12-17T23:59:59.000Z

47

Utilization of the Philippine Research Reactor as a training facility for nuclear power plant operators  

SciTech Connect

The Philippines has a 1-MW swimming-pool reactor facility operated by the Philippine Atomic Energy Commission (PAEC). The reactor is light-water moderated and cooled, graphite reflected, and fueled with 90% enriched uranium. Since it became critical in 1963 it has been utilized for research, radioisotope production, and training. It was used initially in the training of PAEC personnel and other research institutions and universities. During the last few years, however, it has played a key role in training personnel for the Philippine Nuclear Power Project (PNPP).

Palabrica, R.J.

1981-01-01T23:59:59.000Z

48

Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1  

SciTech Connect

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

NONE

1995-08-01T23:59:59.000Z

49

A study on reactor core failure thresholds to safety operation of LMFBR  

SciTech Connect

Japan Nuclear Safety Organization (JNES) has been developing the methodology and computer codes for applying level-1 PSA to LMFBR. Many of our efforts have been directed to the judging conditions of reactor core damage and the time allowed to initiate the accident management. Several candidates of the reactor core failure threshold were examined to a typical proto-type LMFBR with MOX fuel based on the plant thermal-hydraulic analyses to the actual progressions leading to the core damage. The results of the present study showed that the judging condition of coolant-boundary integrity failure, 750 degree-C of the boundary temperature, is enough as the threshold of core damage to PLOHS (protected loss-of-heat sink). High-temperature fuel cladding creep failure will not take place before the coolant-boundary reaches the judging temperature and sodium boiling will not occur due to the system pressure rise. In cases of ATWS (anticipated transient without scrum) the accident progression is so fast and the reactor core damage will be inevitable even a realistic negative reactivity insertion due to the temperature rise is considered. Only in the case of ULOHS (unprotected loss-of-heat sink) a relatively long time of 11 min will be allowed till the shut-down of the reactor before the core damage. (authors)

Kazuo, Haga; Hiroshi, Endo; Tomoko, Ishizu; Yoshihisa, Shindo [Japan Nuclear Energy Safety Organization, Safety Analysis and Evaluation Division, Kamiya-cho MT Bldg., 4-3-20, Toranomon, Minato-ku, Tokyo (Japan)

2006-07-01T23:59:59.000Z

50

Licensed operating reactors: Status summary report, data as of December 31, 1995. Volume 20  

SciTech Connect

The US Nuclear Regulatory Commission`s monthly summary of licensed nuclear power reactor data is based primarily on the operating data report submitted by licensees for each unit. This report is divided into two sections: the first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 availability factors, capacity factors, and forced outage rates are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensees and notes to the use of weighted averages and starting dates other than commercial operation are provided.

NONE

1996-06-01T23:59:59.000Z

51

Load following capability of CANDLE reactor by adjusting coolant operation condition  

SciTech Connect

The load following capability of CANDLE reactor is investigated in the condition that the control rods are unavailable. Both sodium cooled metallic fuel fast reactor (SFR) and {sup 208}Pb cooled metallic fuel fast reactor (LFR) are investigated for their performance in power rate changing by changing its coolant operation condition; either coolant flow rate or coolant inlet temperature. The change by coolant flow rate is difficult especially for SFR because the maximum temperature criteria on cladding material may be violated. The power rate can be changed for its full range easily by changing the coolant temperature at the core inlet. LFR can reduce the same amount of power rate by smaller change of temperature than SFR. However, the coolant output temperature is generally decreased for this method and the thermal efficiency becomes worse.

Sekimoto, Hiroshi; Nakayama, Sinsuke [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology 2-12-1-N1-17, Ookayama, Meguro-ku 152-8550 (Japan)

2012-06-06T23:59:59.000Z

52

Selected Hanford reactor and separations operating data for 1960--1964. Hanford Environmental Dose Reconstruction Project  

SciTech Connect

The purpose of this letter report is to reconstruct from available information that data which can be used to develop daily reactor operating history for 1960--1964. The information needed for source team calculations (as determined by the Source Terms Task Leader) were extracted and included in this report. The data on the amount of uranium dissolved by the separations plants (expressed both as tons and as MW) is also included in this compilation.

Gydesen, S.P.

1992-09-01T23:59:59.000Z

53

Selected Hanford reactor and separations operating data for 1960--1964  

SciTech Connect

The purpose of this letter report is to reconstruct from available information that data which can be used to develop daily reactor operating history for 1960--1964. The information needed for source team calculations (as determined by the Source Terms Task Leader) were extracted and included in this report. The data on the amount of uranium dissolved by the separations plants (expressed both as tons and as MW) is also included in this compilation.

Gydesen, S.P.

1992-09-01T23:59:59.000Z

54

Licensed operating reactors. Status summary report data as of December 31, 1993  

SciTech Connect

The Nuclear Regulatory Commissions annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December, the year to date (in this case calendar year 1993) and cumulative data, usually for the date of commercial operation. The data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided.

Hartfield, R.A.

1994-03-01T23:59:59.000Z

55

Licensed operating reactors: Status summary report data as of December 31, 1991. Volume 16  

SciTech Connect

The Nuclear Regulatory Commission`s annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December because that report contains data for the month of December, the year to date (in this case calendar year 1991) and cumulative data, usually from the date of commercial operation. The data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided.

NONE

1992-03-01T23:59:59.000Z

56

Experiment operations plan for the TH-2 experiment in the NRU reactor. [PWR; BWR  

SciTech Connect

A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the formal operations plan for TH-2--the second experiment in the series of thermal-hydraulic tests conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. The major objective of TH-2 was to develop the experiment reflood control parameters and the procedures to be used in subsequent experiments in this program. In this experiment, the data acquisition and control system was used to control the fuel cladding temperature during a simulated LOCA by using variable reflood coolant flow.

Russcher, G.E.; Wilson, C.L.; Parchen, L.J.; Freshley, M.D.

1983-06-01T23:59:59.000Z

57

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

SciTech Connect

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

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

2010-10-01T23:59:59.000Z

58

Experience with the operation, maintenance and utilisation of the 3 MW TRIGA Mark-II research reactor of Bangladesh  

Science Journals Connector (OSTI)

The 3 MW TRIGA (Training, Research, Isotope, General Atomics) Mark-II research reactor of the Bangladesh Atomic Energy Commission (BAEC) has been operating at Atomic Energy Research Establishment (AERE), Savar, Dhaka, since September 1986. Since its commissioning, the reactor has been used in various fields of research and utilisation, such as Neutron Activation Analysis (NAA), Neutron Radiography (NRG), Neutron Scattering (NS), manpower training and education, and production of radioisotopes for medical applications. The reactor facility encountered a couple of incidents, which were successfully handled by BAEC personnel. In some cases, the help of experts from various local organisations/institutions as well as from the International Atomic Energy Agency (IAEA) was obtained. The upgrading of the Safety Analysis Report (SAR) of the reactor facility was completed in 2005 as per the format of the IAEA Safety Guide, SG-35-G1. The cooling system of the reactor as well as some parts of the instrumentations used in the reactor systems were also upgraded/modified during this period. The paper highlights the experience with the operation, maintenance and utilisation of the research reactor for the last 21 years. It also presents some of the modification and upgrading works carried out to enhance the operational safety of the research reactor.

M.A. Zulquarnain; M.M. Haque; M.A. Salam; M.S. Islam; P.K. Saha; M.A. Sarder; A. Haque; M.A.M. Soner; M.M. Uddin; M.M. Rahman; I. Kamal; M.N. Islam; S.M. Hossain

2009-01-01T23:59:59.000Z

59

Office of Analysis and Evaluation of Operational Data 1989 annual report, Power reactors  

SciTech Connect

The annual report of the US Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) is devoted to the activities performed during 1989. The report is published in two separate parts. This document, NUREG-1272, Vol. 4, No. 1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about the trends of some key performance measures. The report also includes the principal findings and issues identified in AEOD studies over the past year and summarizes information from such sources as licensee event reports, diagnostic evaluations, and reports to the NRC's Operations Center. This report also compiles the status of staff actions resulting from previous Incident Investigation Team (IIT) reports. 16 figs., 9 tabs.

None

1990-07-01T23:59:59.000Z

60

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN C-REACTOR DISASSEMBLY BASIN  

SciTech Connect

C-reactor disassembly basin is being prepared for deactivation and decommissioning (D and D). D and D activities will consist primarily of immobilizing contaminated scrap components and structures in a grout-like formulation. The disassembly basin will be the first area of the C-reactor building that will be immobilized. The scrap components contain aluminum alloy materials. Any aluminum will corrode very rapidly when it comes in contact with the very alkaline grout (pH > 13), and as a result would produce hydrogen gas. To address this potential deflagration/explosion hazard, Savannah River National Laboratory (SRNL) reviewed and evaluated existing experimental and analytical studies of this issue to determine if any process constraints are necessary. The risk of accumulation of a flammable mixture of hydrogen above the surface of the water during the injection of grout into the C-reactor disassembly area is low if the assessment of the aluminum surface area is reliable. Conservative calculations estimate that there is insufficient aluminum present in the basin areas to result in significant hydrogen accumulation in this local region. The minimum safety margin (or factor) on a 60% LFL criterion for a local region of the basin (i.e., Horizontal Tube Storage) was greater than 3. Calculations also demonstrated that a flammable situation in the vapor space above the basin is unlikely. Although these calculations are conservative, there are some measures that may be taken to further minimize the risk of developing a flammable condition during grouting operations.

Wiersma, B.

2011-07-12T23:59:59.000Z

Note: This page contains sample records for the topic "reactor operators typically" 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

Study on the operational control of helical fusion reactor FFHR-d1  

Science Journals Connector (OSTI)

Abstract Plasma operation control of the Large Helical Device (LHD)-type helical fusion reactor, FFHR-d1, was examined by using a quasi-one-dimensional particle balance calculation code. It was found that feedback control of the fusion power, which is proposed in the past study with a zero-dimensional model, was difficult because of the time delay between the pellet injection and response of the fusion power due to shallow pellet deposition. While the achievable parameter range of the fusion power at a steady state is limited by the pellet injection condition and magnetic field strength, smooth ignition access and steady-state sustainment can in principle be achieved via feedback control of the line-averaged electron density by manipulating the pellet fuelling rate and maintaining adequate control of the external heating power.

Takuya Goto; Junichi Miyazawa; Ryuichi Sakamoto; Osamu Mitarai; Akio Sagara

2014-01-01T23:59:59.000Z

62

Office for Analysis and Evaluation of Operational Data 1996 annual report. Volume 10, Number 1: Reactors  

SciTech Connect

This annual report of the US Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) describes activities conducted during 1996. The report is published in three parts. NUREG-1272, Vol. 10, No. 1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about trends of some key performance measures. The report also includes the principal findings and issues identified in AEOD studies over the past year and summarizes information from such sources as licensee event reports and reports to the NRC`s Operations Center. NUREG-1272, Vol. 10, No. 2, covers nuclear materials and presents a review of the events and concerns during 1996 associated with the use of licensed material in nonreactor applications, such as personnel overexposures and medical misadministrations. Both reports also contain a discussion of the Incident Investigation Team program and summarize both the Incident Investigation Team and Augmented Inspection Team reports. Each volume contains a list of the AEOD reports issued from CY 1980 through 1996. NUREG-1272, Vol. 10, No. 3, covers technical training and presents the activities of the Technical Training Center in support of the NRC`s mission in 1996.

NONE

1997-12-01T23:59:59.000Z

63

First operation with the JET International Thermonuclear Experimental Reactor-like wall  

SciTech Connect

To consolidate International Thermonuclear Experimental Reactor (ITER) design choices and prepare for its operation, Joint European Torus (JET) has implemented ITER's plasma facing materials, namely, Be for the main wall and W in the divertor. In addition, protection systems, diagnostics, and the vertical stability control were upgraded and the heating capability of the neutral beams was increased to over 30 MW. First results confirm the expected benefits and the limitations of all metal plasma facing components (PFCs) but also yield understanding of operational issues directly relating to ITER. H-retention is lower by at least a factor of 10 in all operational scenarios compared to that with C PFCs. The lower C content (? factor 10) has led to much lower radiation during the plasma burn-through phase eliminating breakdown failures. Similarly, the intrinsic radiation observed during disruptions is very low, leading to high power loads and to a slow current quench. Massive gas injection using a D{sub 2}/Ar mixture restores levels of radiation and vessel forces similar to those of mitigated disruptions with the C wall. Dedicated L-H transition experiments indicate a 30% power threshold reduction, a distinct minimum density, and a pronounced shape dependence. The L-mode density limit was found to be up to 30% higher than for C allowing stable detached divertor operation over a larger density range. Stable H-modes as well as the hybrid scenario could be re-established only when using gas puff levels of a few 10{sup 21} es{sup ?1}. On average, the confinement is lower with the new PFCs, but nevertheless, H factors up to 1 (H-Mode) and 1.3 (at ?{sub N}?3, hybrids) have been achieved with W concentrations well below the maximum acceptable level.

Neu, R. [EFDA-CSU, Boltzmannstr. 2, 85748 Garching (Germany) [EFDA-CSU, Boltzmannstr. 2, 85748 Garching (Germany); Max-Planck-Institut für Plasmaphysik, Euratom Association, Boltzmannstr. 2, 85748 Garching (Germany); Arnoux, G.; Beurskens, M.; Challis, C.; Giroud, C.; Lomas, P.; Maddison, G.; Matthews, G.; Mayoral, M.-L.; Meigs, A.; Rimini, F. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)] [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Bobkov, V.; Dux, R.; Hobirk, J.; Lang, P.; Maggi, C.; Pütterich, T.; Sertoli, M.; Sieglin, B. [Max-Planck-Institut für Plasmaphysik, Euratom Association, Boltzmannstr. 2, 85748 Garching (Germany)] [Max-Planck-Institut für Plasmaphysik, Euratom Association, Boltzmannstr. 2, 85748 Garching (Germany); Brezinsek, S. [IEK-4, Association EURATOM/Forschungszentrum Jülich GmbH, Jülich 52425 (Germany)] [IEK-4, Association EURATOM/Forschungszentrum Jülich GmbH, Jülich 52425 (Germany); and others

2013-05-15T23:59:59.000Z

64

EIS-0147: Continued Operation of the K-,L-, and P- Reactors, Savannah River Site, Aiken, South Carolina  

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

This environmental impact statement (EIS) analyzes the environmental impacts of the proposed action, which is to continue operation of K-, L-, and P-Reactors at the Savannah River Site (SRS) to ensure the capability to produce nuclear materials, and to produce nuclear materials as necessary for United States defense and nondefense programs.

65

Statistical analysis of operational reliability for electric pump units CN 60–180 of VVER-1000 reactors with root estimation methods  

Science Journals Connector (OSTI)

We consider the problem of processing statistical data obtained during the operation of pumping units CN 60–180 operating as standard equipment of the VVER-1000 reactors. Data for analysis was taken from the indu...

A. V. Antonov; V. A. Chepurko; N. G. Zyulyaeva…

2010-07-01T23:59:59.000Z

66

Computer analyses for the design, operation and safety of new isotope production reactors: A technology status review  

SciTech Connect

A review is presented on the currently available technologies for nuclear reactor analyses by computer. The important distinction is made between traditional computer calculation and advanced computer simulation. Simulation needs are defined to support the design, operation, maintenance and safety of isotope production reactors. Existing methods of computer analyses are categorized in accordance with the type of computer involved in their execution: micro, mini, mainframe and supercomputers. Both general and special-purpose computers are discussed. Major computer codes are described, with regard for their use in analyzing isotope production reactors. It has been determined in this review that conventional systems codes (TRAC, RELAP5, RETRAN, etc.) cannot meet four essential conditions for viable reactor simulation: simulation fidelity, on-line interactive operation with convenient graphics, high simulation speed, and at low cost. These conditions can be met by special-purpose computers (such as the AD100 of ADI), which are specifically designed for high-speed simulation of complex systems. The greatest shortcoming of existing systems codes (TRAC, RELAP5) is their mismatch between very high computational efforts and low simulation fidelity. The drift flux formulation (HIPA) is the viable alternative to the complicated two-fluid model. No existing computer code has the capability of accommodating all important processes in the core geometry of isotope production reactors. Experiments are needed (heat transfer measurements) to provide necessary correlations. It is important for the nuclear community, both in government, industry and universities, to begin to take advantage of modern simulation technologies and equipment. 41 refs.

Wulff, W.

1990-01-01T23:59:59.000Z

67

The determination of neutron flux in the Texas A & M triga reactor during pulse and steady-state operations  

E-Print Network (OSTI)

-state operation. Neutron flux measurement during a pulse presents an additional problem in that the flux levels vary during the rapid rise and fall in reactor power. The power level transient of the reactor was followed, using the current output of a boron-10... as a flux monitor only at low power. levels or neutron fluxes. The antimony flux monitor in the steady-state flux measurement showed the same type of increase in flux magni- tude as that of the pulse measurement. The high flux values at steady...

O'Donnell, John Joseph

2012-06-07T23:59:59.000Z

68

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R- AND P-REACTOR VESSELS  

SciTech Connect

The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS&T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D&D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or Portland cement grout (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters does not provide a significant margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. It is recommended that this grout not be utilized for this task. The R-reactor vessel contains significantly less aluminum and thus a Portland cement grout may be considered as well. For example, if the grout fill rate is less than 1 inch/min and the grout temperature is maintained at 70 C or less, the risk of hydrogen accumulation in the R-reactor vessel is very low for the Portland cement. Alternatively, if the grout fill rate is less than 0.5 inch/min and the grout is maintained at a temperature of 80 C, the risk will again be very low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. For P-reactor, grout temperatures less than 100 C should provide an adequate safety margin for the pH 8 and pH 10.4 grout formulations. For R-reactor, grout temperatures less than 70 C or 80 C will provide an adequate safety margin for the Portland cement. The other grout formulations are also viable options for R-reactor. (2) Minimize the grout fill rate as much as practical. Lowering the fill rate takes advantage of passivation of the aluminum components and hence lower hydrogen generation rates. For P-reactor, fill rates that are less than 2 inches/min for the ceramicrete and the silica fume grouts will reduce the chance of significant hydrogen accumulation. For R-reactor, fill rates less than 1 inch/min will again minimize the risk of hydrogen accumulation. (3) Ventilate the building as much as practical (e.g., leave doors open) to further disperse hydrogen. The volumetric hydrogen generation rates in the P-reactor vessel, however, are low for the pH 8 and pH 10.4 grout, (i.e., less than 0.32 ft{sup 3}/min). If further walk-down inspections of the reactor vessels suggest an increase in the actual areal density of aluminum, the calculations should be re-visited.

Wiersma, B.

2009-12-29T23:59:59.000Z

69

Operation of a steam hydro-gasifier in a fluidized bed reactor  

E-Print Network (OSTI)

Using Self-Sustained Hydro- Gasification." [0011] In aprocess, using a steam hydro-gasification reactor (SHR) thepyrolysis and hydro-gasification in a single step. This

Park, Chan Seung; Norbeck, Joseph N.

2008-01-01T23:59:59.000Z

70

MTG fluidized bed reactor–regenerator unit with catalyst circulation: process simulation and operation of an experimental setup  

Science Journals Connector (OSTI)

Simulation of the MTG process carried out in a fluidized bed reactor–regenerator system with catalyst circulation is studied by using the kinetic results obtained in an experimental unit. Data considered covered a wide range of operating conditions including temperature, space time and average residence times in both reactor and regenerator. This simulation is based on the use of adequate kinetic equations for the main MTG reaction, for the catalyst deactivation and for catalyst regeneration. In addition, a third stage in the process allowing for coke equilibration, prior to its combustion, is also included. Catalyst loss due to attrition has also been taken into account. An objective function based on relative production rate is optimized by changing systematically the process parameters such as temperature, space time and catalyst activity. Results are also validated in the experimental unit and demonstrate the simplicity of the reactor–regenerator system with catalyst circulation and the versatility of this configuration for carrying out the MTG process.

Ana G. Gayubo; Jose M. Ortega; Andres T. Aguayo; Jose M. Arandes; Javier Bilbao

2000-01-01T23:59:59.000Z

71

Operation of a steam hydro-gasifier in a fluidized bed reactor  

E-Print Network (OSTI)

OF A S T E A M HYDRO-GASIFIER IN A FLUIDIZED BED REACTOROF A S T E A M HYDRO-GASIFIER IN A FLUIDIZED BED REACTOR F Iis fed into a hydro-gasifier reactor. One such process was

Park, Chan Seung; Norbeck, Joseph N.

2008-01-01T23:59:59.000Z

72

Sliding Mode Control for Pressurized-Water Nuclear Reactors in load following operations with bounded xenon oscillations  

Science Journals Connector (OSTI)

Abstract One of the important operations in nuclear power plants is load-following in which imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation considered to be a constraint for the load-following operation. In this paper, sliding mode control (SMC) which is a robust nonlinear controller is designed to control the Pressurized-Water Nuclear Reactor (PWR) power for the load-following operation problem that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO) strategy to maintain xenon oscillations to be bounded. The constant AO is a robust state constraint for load-following problem. The reactor core is simulated based on the two-point nuclear reactor model and one delayed neutron group. The stability analysis is given by means Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability. Results show that the proposed controller for the load-following operation is sufficiently effective so that the xenon oscillations are kept bounded in the considered region.

G.R. Ansarifar; S. Saadatzi

2015-01-01T23:59:59.000Z

73

EXPERIMENT OPERATIONS PLAN FOR A LOSS-OF-COOLANT ACCIDENT SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR  

SciTech Connect

Pressurized water reactor loss-of-coolant accident phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship between the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. This document contains both experiment proposal and assembly proposal information. The intent of this document is to supply information required by the Chalk River Nuclear Laboratories (CRNL), and to identify the planned procedures and data that will be used both to establish readiness to proceed from one test phase to the next and to operate the experiment. Operating control settings and limits are provided for both experimenter systems and CRNL systems. A hazards review summarizes safety issues that have been addressed during the development of the experiment plan.

Russcher, G. E.; Cannon, L. W.; Goodman, R. L.; Hesson, G. M.; King, L. L.; McDuffie, P. N.; Marshall, R. K.; Nealley, C.; Pilger, J. P.; Mohr, C. L.

1981-04-01T23:59:59.000Z

74

Blue Ribbon Commission, Yucca Mountain Closure, Court Actions - Future of Decommissioned Reactors, Operating Reactors and Nuclear Power - 13249  

SciTech Connect

Issues related to back-end of the nuclear fuel cycle continue to be difficult for the commercial nuclear power industry and for the decision makers at the national and international level. In the US, the 1982 NWPA required DOE to develop geological repositories for SNF and HLW but in spite of extensive site characterization efforts and over ten billion dollars spent, a repository opening is nowhere in sight. There has been constant litigation against the DOE by the nuclear utilities for breach of the 'standard contract' they signed with the DOE under the NWPA. The SNF inventory continues to rise both in the US and globally and the nuclear industry has turned to dry storage facilities at reactor locations. In US, the Blue Ribbon Commission on America's Nuclear Future issued its report in January 2012 and among other items, it recommends a new, consent-based approach to siting of facilities, prompt efforts to develop one or more geologic disposal facilities, and prompt efforts to develop one or more consolidated storage facilities. In addition, the March 2011 Fukushima Daiichi accident had a severe impact on the future growth of nuclear power. The nuclear industry is focusing on mitigation strategies for beyond design basis events and in the US, the industry is in the process of implementing the recommendations from NRC's Near Term Task Force. (authors)

Devgun, Jas S. [Nuclear Power Technologies, Sargent and Lundy LLC1, Chicago, IL (United States)] [Nuclear Power Technologies, Sargent and Lundy LLC1, Chicago, IL (United States)

2013-07-01T23:59:59.000Z

75

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

76

Role of research reactors in training of NPP personnel with special focus on training reactor VR-1  

SciTech Connect

Research reactors play an important role in providing key personnel of nuclear power plants a hands-on experience from operation and experiments at nuclear facilities. Training of NPP (Nuclear Power Plant) staff is usually deeply theoretical with an extensive utilisation of simulators and computer visualisation. But a direct sensing of the reactor response to various actions can only improve the personnel awareness of important aspects of reactor operation. Training Reactor VR-1 and its utilization for training of NPP operators and other professionals from Czech Republic and Slovakia is described. Typical experimental exercises and good practices in organization of a training program are demonstrated. (authors)

Sklenka, L.; Rataj, J.; Frybort, J.; Huml, O. [Dept. of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical Univ. in Prague, V Holesovickach 2, Prague 8, 180 00 (Czech Republic)

2012-07-01T23:59:59.000Z

77

Photo of the Week: Not Your Typical Jet Engine | Department of Energy  

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

Not Your Typical Jet Engine Not Your Typical Jet Engine Photo of the Week: Not Your Typical Jet Engine November 23, 2012 - 11:57am Addthis As part of the Aircraft Nuclear Propulsion Program, the U.S. conducted extensive research showing that nuclear fission could power an aircraft. The research involved a series of Heat Transfer Reactor Experiments (HTREs), which tested if different types of jet engines could be run by nuclear power. In 1955, however, the project was cancelled, and a safe, operational prototype aircraft was never developed. In this 1988 photo, the two HTRE reactors are shown in transport to Idaho National Laboratory's EBR-1 visitor center, where they remain today. | Photo courtesy of Idaho National Laboratory. As part of the Aircraft Nuclear Propulsion Program, the U.S. conducted

78

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS  

SciTech Connect

The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D and D). D and D activities consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS and T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D and D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or Portland cement groupt (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters does not provide a margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. It is recommended that this grout not be utilized for this task. The R-reactor vessel cotnains significantly less aluminum based on current facility process knowledge, surface observations, and drawings. Therefore, a Portland cement grout may be considered for grouting operations as well as the other grout formulations. For example, if the grout fill rate is less than 1 inch/min and the grout temperature is maintained at 70 C or less, the risk of hydrogen accumulation during fill operations in the R-reactor vessel is low for the Portland cement. Alternatively, if the grout fill rate is less than 0.5 inch/min and the grout is maintained at a temperature of 80 C, the risk is again low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. For P-reactor, grout temperatures less than 100 C should provide an adequate safety margin for the pH 8 and pH 10.4 grout formulations. For R-reactor, grout temperatures less than 70 C or 80 C will provide an adequate safety margin for the Portland cement. The other grout formulations are also viable options for R-reactor. (2) Minimize the grout fill rate as much as practical. Lowering the fill rate takes advantage of passivation of the aluminum components and hence lower hydrogen generation rates. For P-reactor, fill rates that are less than 2 inches/min for the ceramicrete and the silica fume grouts will reduce the chance of significant hydrogen accumulation. For R-reactor, fill rates less than 1 inch/min will again minimize the risk of hydrogen accumulation. (3) Ventilate the building as much as practical (e.g., leave doors open) to further disperse hydrogen. The volumetric hydrogen generation rates in the P-reactor vessel, however, are low for the pH 8 and pH 10.4 grout, (i.e., less than 0.97 ft{sup 3}/min). If further walk-down inspections of the reactor vessels suggest an increase in the actual areal density of aluminum, the calculations should be re-visited.

Wiersma, B.

2010-05-24T23:59:59.000Z

79

Record of Cycling Operation of the Natural Nuclear Reactor in the Oklo/Okelobondo Area in Gabon  

Science Journals Connector (OSTI)

Using selective laser extraction technique combined with sensitive ion-counting mass spectrometry, we have analyzed the isotopic structure of fission noble gases in U-free La-Ce-Sr-Ca aluminous hydroxy phosphate associated with the 2 billion yr old Oklo natural nuclear reactor. In addition to elevated abundances of fission-produced Zr, Ce, and Sr, we discovered high (up to 0.03??cm3???STP/g) concentrations of fission Xe and Kr, the largest ever observed in any natural material. The specific isotopic structure of xenon in this mineral defines a cycling operation for the reactor with 30-min active pulses separated by 2.5 h dormant periods. Thus, nature not only created conditions for self-sustained nuclear chain reactions, but also provided clues on how to retain nuclear wastes, including fission Xe and Kr, and prevent uncontrolled runaway chain reaction.

A. P. Meshik; C. M. Hohenberg; O. V. Pravdivtseva

2004-10-27T23:59:59.000Z

80

TMRBAR: a code to calculate plasma parameters for tandem-mirror reactors operating in the MARS mode  

SciTech Connect

The purpose of this report is to document the plasma power balance model currently used by LLNL to calculate steady state operating points for tandem mirror reactors. The code developed from this model, TMRBAR, has been used to predict the performance and define supplementary heating requirements for drivers used in the Mirror Advanced Reactor Study (MARS) and for the Fusion Power Demonstration (FPD) study. The equations solved included particle and energy balance for central cell and end cell species, quasineutrality at several cardinal points in the end cell region, as well as calculations of volumes, densities and average energies based on given constraints of beta profiles and fusion power output. Alpha particle ash is treated self-consistently, but no other impurity species is treated.

Campbell, R.B.

1983-08-30T23:59:59.000Z

Note: This page contains sample records for the topic "reactor operators typically" 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

Report to the US Nuclear Regulatory Commission on analysis and evaluation of operational data - 1987: Power reactors  

SciTech Connect

This annual report of the US Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) is devoted to the activities performed during 1987. The report is published in two volumes. NUREG-1272, Vol. 2, No. 1, covers Power Reactors and presents an overview of the operating experience of the nuclear power industry, with comments regarding the trends of some key performance measures. The report also includes the principal findings and issues identified in AEOD studies over the past year, and summarizes information from Licensee Event Reports, the NRC's Operations Center, and Diagnostic Evaluations. NUREG-1272, Vol. 2, No. 2, covers Nonreactors and presents a review of the nonreactors events and misadministration reports that were reported in 1987 and a brief synopsis of AEOD studies published in 1987. Each volume contains a list of the AEOD Reports issued for 1980-1987.

none,

1988-10-01T23:59:59.000Z

82

Research and Development of High Temperature Light Water Cooled Reactor Operating at Supercritical-Pressure in Japan  

SciTech Connect

This paper summarizes the status and future plans of research and development of the high temperature light water cooled reactor operating at supercritical-pressure in Japan. It includes; the concept development; material for the fuel cladding; water chemistry under supercritical pressure; thermal hydraulics of supercritical fluid; and the conceptual design of core and plant system. Elements of concept development of the once-through coolant cycle reactor are described, which consists of fuel, core, reactor and plant system, stability and safety. Material studies include corrosion tests with supercritical water loops and simulated irradiation tests using a high-energy transmission electron microscope. Possibilities of oxide dispersion strengthening steels for the cladding material are studied. The water chemistry research includes radiolysis and kinetics of supercritical pressure water, influence of radiolysis and radiation damage on corrosion and behavior on the interface between water and material. The thermal hydraulic research includes heat transfer tests of single tube, single rod and three-rod bundles with a supercritical Freon loop and numerical simulations. The conceptual designs include core design with a three-dimensional core simulator and sub-channel analysis, and balance of plant. (authors)

Yoshiaki Oka [Nuclear Engineering Research Laboratory, The University of Tokyo, 7-3-1, Hongo, Bunkyo-ku, Tokyo, 112-0006 (Japan); Katsumi Yamada [Isogo Nuclear Engineering Center, Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan)

2004-07-01T23:59:59.000Z

83

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS  

SciTech Connect

The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS&T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D&D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or portland cement grout (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the R- and P- reactor vessels. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Conservative calculations estimate that either ceramicrete or the silica fume grout may be used to safely grout the R- and P- reactor vessels. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. Grout temperatures less than 100 C should however, still provide an adequate safety margin for the pH 8 and pH 10.4 grout formulations. (2) Minimize the fill rate as much as practical. Lowering the fill rate takes advantage of passivation of the aluminum components and hence lower hydrogen generation rates. Fill rates that are less than 2 inches/min will reduce the chance of significant hydrogen build-up. (3) Ventilate the building as much as practical (e.g., leave doors open) to further disperse hydrogen. The volumetric hydrogen generation rates however, are low for the pH 8 and pH 10.4 grout, i.e., less than 0.32 ft{sup 3}/min. Portland cement grout, on the other hand, for the same range of process parameters does not provide a significant margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations. It is recommended that this grout not be utilized for this task. If further walk-down inspections of the reactor vessels suggest an increase in the actual areal density of aluminum, the calculations should be re-visited.

Wiersma, B.

2009-10-29T23:59:59.000Z

84

The development and operational testing of an experimental reactor for gas-liquid-solid reaction systems at high temperatures and pressures  

E-Print Network (OSTI)

shaft. With the impeller in place and rotating, gas was drawn into the top port and ejected at the impeller mount. The reactor pressure was monitored via the transducer port. The transducer was a Viatran Pressure Transducer, model 103. The liquid...THE DEVELOPMENT AND OPERATIONAL TESTING OF AN EXPERIMENTAL REACTOR FOR GAS-LIQUID-SOLID REACTION SYSTEMS AT HIGH TEMPERATURES AND PRESSURES A Thesis by RICHARD KENNETH HESS Submitted to the Graduate College of Texas A&M University in partial...

Hess, Richard Kenneth

2012-06-07T23:59:59.000Z

85

Removal of 1,082-Ton Reactor Among Richland Operations Office’s 2014 Accomplishments  

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

RICHLAND, Wash. – Workers with EM’s Richland Operations Office and its contractors made progress this year in several areas of Hanford site cleanup that helped protect employees, the public, environment, and Columbia River.

86

Operating experience feedback report -- turbine-generator overspeed protection systems: Commercial power reactors. Volume 11  

SciTech Connect

This report presents the results of the US Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) review of operating experience of main turbine-generator overspeed and overspeed protection systems. It includes an indepth examination of the turbine overspeed event which occurred on November 9, 1991, at the Salem Unit 2 Nuclear Power Plant. It also provides information concerning actions taken by other utilities and the turbine manufacturers as a result of the Salem overspeed event. AEOD`s study reviewed operating procedures and plant practices. It noted differences between turbine manufacturer designs and recommendations for operations, maintenance, and testing, and also identified significant variations in the manner that individual plants maintain and test their turbine overspeed protection systems. AEOD`s study provides insight into the shortcomings in the design, operation, maintenance, testing, and human factors associated with turbine overspeed protection systems. Operating experience indicates that the frequency of turbine overspeed events is higher than previously thought and that the bases for demonstrating compliance with NRC`s General Design Criterion (GDC) 4, Environmental and dynamic effects design bases, may be nonconservative with respect to the assumed frequency.

Ornstein, H.L.

1995-04-01T23:59:59.000Z

87

Computerized operating procedures for shearing and dissolution of segments from LWBR (Light Water Breeder Reactor) fuel rods  

SciTech Connect

This report presents two detailed computerized operating procedures developed to assist and control the shearing and dissolution of irradiated fuel rods. The procedures were employed in the destructive analysis of end-of-life fuel rods from the Light Water Breeder Reactor (LWBR) that was designed by the Westinghouse Electric Corporation Bettis Atomic Power Laboratory. Seventeen entire fuel rods from the end-of-life core of the LWBR were sheared into 169 precisely characterized segments, and more than 150 of these segments were dissolved during execution of the LWBR Proof-of-Breeding (LWBR-POB) Analytical Support Project at Argonne National Laboratory. The procedures illustrate our approaches to process monitoring, data reduction, and quality assurance during the LWBR-POB work.

Osudar, J.; Deeken, P.G.; Graczyk, D.G.; Fagan, J.E.; Martino, F.J.; Parks, J.E.; Levitz, N.M.; Kessie, R.W.; Leddin, J.M.

1987-05-01T23:59:59.000Z

88

Office for Analysis and Evaluation of Operational Data 1992 annual report: Power reactors. Volume 7, No. 1  

SciTech Connect

The annual report of the US Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) is devoted to the activities performed during 1992. The report is published in two separate parts. NUREG-1272, Vol. 7, No. 1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about the trends of some key performance, measures. The report also includes the principal findings and issues identified in AEOD studies over the past year, and summarizes information from such sources as licensee event report% diagnostic evaluations, and reports to the NRC`s Operations Center. The reports contain a discussion of the Incident Investigation Team program and summarize the Incident Investigation Team and Augmented Inspection Team reports for that group of licensees. NUREG-1272, Vol. 7, No. 2, covers nonreactors and presents a review of the events and concerns during 1992 associated with the use of licensed material in nonreactor applications, such as personnel overexposures and medical misadministrations. Each volume contains a list of the AEOD reports issued for 1984--1992.

none,

1993-07-01T23:59:59.000Z

89

Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Test 2  

SciTech Connect

A loss-of-coolant accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects on pressurized water reactor (PWR) test fuel bundles. This Experiment Operation Plan (EOP) Addendum 2, together with the referenced EOP, describes the desired operating conditions and additional hazards review associated with the four-part MT-2 experiment. The primary portions of the experiment, MT-2.2 and MT-2.3, will evaluate the following: 1) the mechanical deformation of pressurized fuel rods subjected to a slow LOCA, using reflood water for temperature control, that is designed to produce cladding temperatures in the range from 1033 to 1089K (1400 to 1500°F) for an extended time, and 2) the effects of the deformed and possibly failed cladding on the thermal-hydraulic performance of the test assembly during simulated LOCA heating and reflooding. The secondary portions of the experiment, MT-2.1 and MT-2.4, are intended to provide thermal-hydraulic calibration information during two-stage reflood conditions for 1) relatively low cladding temperatures, <839K (1050°F), on nondeformed rods, and 2) moderately high cladding temperatures, <1089K (1500°F), on deformed rods.

Russcher, G. E.; Barner, J. O.; Hesson, G. M.; Wilson, C. L.; Parchen, L. J.; Cunningham, M. E.; Marshall, R. K.; Mohr, C. L.

1981-09-01T23:59:59.000Z

90

naval reactors  

National Nuclear Security Administration (NNSA)

After operating for 34 years and training over 14,000 sailors, the Department of Energy S1C Prototype Reactor Site in Windsor, Connecticut, was returned to "green field"...

91

Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Tests 1 and 2  

SciTech Connect

A loss of Coolant Accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects of LOCA conditions on pressurized water reactor test fuel bundles. This experiment operation plan for the second and third experiments of the program will provide peak fuel cladding temperatures of up to 1172K (1650{degree}F) and 1061K (1450{degree}) respectively. for a long enough time to cause test fuel cladding deformation and rupture in both. Reflood coolant delay times and the reflooding rates for the experiments were selected from thermal-hydraulic data measured in the National Research Universal (NRU) reactor facilities and test train assembly during the first experiment.

Russcher, G. E.; Wilson, C. L.; Marshall, R, K.; King, L. L.; Parchen, L. J.; Pilger, J. P.; Hesson, G. M.; Mohr, C. L.

1981-09-01T23:59:59.000Z

92

Hydroelectric power provides a cheap source of electricity with few carbon emissions. Yet, reservoirs are not operated sustainably, which we define as meeting societal needs for water and power while protecting long-term health of the river ecosystem. Reservoirs that generate hydropower are typically operated with the goal of maximizing energy reve  

SciTech Connect

Hydroelectric power provides a cheap source of electricity with few carbon emissions. Yet, reservoirs are not operated sustainably, which we define as meeting societal needs for water and power while protecting long-term health of the river ecosystem. Reservoirs that generate hydropower are typically operated with the goal of maximizing energy revenue, while meeting other legal water requirements. Reservoir optimization schemes used in practice do not seek flow regimes that maximize aquatic ecosystem health. Here, we review optimization studies that considered environmental goals in one of three approaches. The first approach seeks flow regimes that maximize hydropower generation, while satisfying legal requirements, including environmental (or minimum) flows. Solutions from this approach are often used in practice to operate hydropower projects. In the second approach, flow releases from a dam are timed to meet water quality constraints on dissolved oxygen (DO), temperature and nutrients. In the third approach, flow releases are timed to improve the health of fish populations. We conclude by suggesting three steps for bringing multi-objective reservoir operation closer to the goal of ecological sustainability: (1) conduct research to identify which features of flow variation are essential for river health and to quantify these relationships, (2) develop valuation methods to assess the total value of river health and (3) develop optimal control softwares that combine water balance modelling with models that predict ecosystem responses to flow.

Jager, Yetta [ORNL; Smith, Brennan T [ORNL

2008-02-01T23:59:59.000Z

93

Assessment of torsatrons as reactors  

SciTech Connect

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R[sub 0] = 6.6-8.8 m, on-axis magnetic field B[sup 0] = 4.8-7.5 T, B[sub max] (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions.

Lyon, J.F. (Oak Ridge National Lab., TN (United States)) [Oak Ridge National Lab., TN (United States); Painter, S.L. (Australian National Univ., Canberra, ACT (Australia)) [Australian National Univ., Canberra, ACT (Australia)

1992-12-01T23:59:59.000Z

94

Assessment of torsatrons as reactors  

SciTech Connect

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R{sub 0} = 6.6-8.8 m, on-axis magnetic field B{sup 0} = 4.8-7.5 T, B{sub max} (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions.

Lyon, J.F. [Oak Ridge National Lab., TN (United States); Painter, S.L. [Australian National Univ., Canberra, ACT (Australia)

1992-12-01T23:59:59.000Z

95

Design and operation of a rotating drum radio frequency plasma reactor for the modification of free nanoparticles  

SciTech Connect

A rotating drum rf plasma reactor was designed to functionalize the surface of nanoparticles and other unusually shaped substrates through plasma polymerization and surface modification. This proof-of-concept reactor design utilizes plasma polymerized allyl alcohol to add OH functionality to Fe{sub 2}O{sub 3} nanoparticles. The reactor design is adaptable to current plasma hardware, eliminating the need for an independent reactor setup. Plasma polymerization performed on Si wafers, Fe{sub 2}O{sub 3} nanoparticles supported on Si wafers, and freely rotating Fe{sub 2}O{sub 3} nanoparticles demonstrated the utility of the reactor for a multitude of processes. X-ray photoelectron spectroscopy and Fourier transform infrared spectroscopy were used to characterize the surface of the substrates prior to and after plasma deposition, and scanning electron microscopy was used to verify that no extensive change in the size or shape of the nanoparticles occurred because of the rotating motion of the reactor. The reactor design was also extended to a non-depositing NH{sub 3} plasma modification system to demonstrate the reactor design is effective for multiple plasma processes.

Shearer, Jeffrey C.; Fisher, Ellen R. [Department of Chemistry, Colorado State University, Fort Collins, Colorado 80523-1872 (United States)

2013-06-15T23:59:59.000Z

96

Study of power distribution in the CZP, HFP and normal operation states of VVER-1000 (Bushehr) nuclear reactor core by coupling nuclear codes  

Science Journals Connector (OSTI)

Abstract In this research, the simulation of one-sixth of VVER-1000 (Bushehr) reactor core is carried out by WIMS-D4 nuclear code, based on symmetry of core and also by information obtained from FSAR. The cross sections of some nuclides are obtained by WIMS-D4 from the beginning of cycle (BOC) to the end of cycle (EOC), and they are transferred into the CITATION code as inputs. In the next stage, the amounts of neutron fluxes and power of reactor core are obtained by CITATION code in the CZP and HFP states. Then, the received products are returned again into the extended program cycle, thereby distributions of neutron fluxes and power are finally depicted. In the meantime, the space distribution of neutron fluxes and power throughout the core are presented during the normal operation by this simulation. It can be inferred that if the reactor operation continues, a flat power distribution will be made in the reactor core that might cause maximum power.

Mohsen Rafiei Karahroudi; Seyed Alireza Mousavi Shirazi

2015-01-01T23:59:59.000Z

97

Gearbox Typical Failure Modes, Detection, and Mitigation Methods (Presentation)  

SciTech Connect

This presentation was given at the AWEA Operations & Maintenance and Safety Seminar and focused on what the typical gearbox failure modes are, how to detect them using detection techniques, and strategies that help mitigate these failures.

Sheng, S.

2014-01-01T23:59:59.000Z

98

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

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

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

99

Nuclear reactors in the United States  

Science Journals Connector (OSTI)

Nuclear reactors in the United States ... A chart listing the operating and planned nuclear reactors in the United States. ... Nuclear / Radiochemistry ...

Hubert N. Alyea

1956-01-01T23:59:59.000Z

100

Granular Dynamics in Pebble Bed Reactor Cores  

E-Print Network (OSTI)

pebble bed reactor,” Nuclear Engineering and Design, vol.the AVR reactor,” Nuclear Engineering and Design, vol. 121,Operating Experience,” Nuclear Engineering and Design, vol.

Laufer, Michael Robert

2013-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor operators typically" 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.


101

Analysis of the antineutrino rate during CANDU reactor startup.  

E-Print Network (OSTI)

??Detection systems used to monitor reactor operations are of significant interest as tools for verification of operator declarations. Current reactor site safeguards are limited to… (more)

Matthews, Christopher

2012-01-01T23:59:59.000Z

102

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

103

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. 1 figure.

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

1994-12-20T23:59:59.000Z

104

Nuclear Reactor (atomic reactor)  

Science Journals Connector (OSTI)

A nuclear reactor splits Uranium or Plutonium nuclei, and the...235 is fissionable but more than 99% of the naturally occurring Uranium is U238 that makes enrichment mandatory. In some reactors U238 and Thorium23...

2008-01-01T23:59:59.000Z

105

Light Water Reactor Sustainability (LWRS) Initiative Science-Based R&D to Extend Nuclear Plant Operation  

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

Nuclear Energy Nuclear Energy Updates Dr. Pete Lyons Acting Assistant Secretary for Nuclear Energy U.S. Department of Energy December 9, 2010 NEAC Meeting Leadership Changes Pete Miller retired Pete Lyons - Acting NE-1 Shane Johnson - Acting NE-2 Dennis Miotla - Acting COO Monica Regalbuto - Acting DAS for Fuel Cycle Technologies John Herczeg- Acting ADAS for Fuel Cycle Technologies John Kelly - DAS for Nuclear Reactor Technologies Bob Boudreau- Acting ADAS International Nuclear Energy Coop Monica Regalbuto John Kelly NE University Programs (NEUP) - Overview and FY 2011 Schedule NEUP FY 2011 Solicitations Schedule RPA/FOA Pre- Applications Proposals Due Awards Announced R&D (PS and Blue Sky) Oct. '10 Dec. '10 Feb. '11 May '11 Integrated Research Projects (IRP) Dec. '10 Late Jan '11

106

Light Water Reactor Sustainability (LWRS) Initiative Science-Based R&D to Extend Nuclear Plant Operation  

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

9, 2010 9, 2010 New Program Proposal for Fiscal Year 2011 - Modified Open Cycle Carter "Buzz" Savage Nuclear Energy Advisory Committee Meeting April 29, 2010 Washington, DC April 29, 2010 Recycle of Used Fuel Option to recycle used fuel has been the subject of much debate and discussion. Nonproliferation issues and economics have limited recycle options. Recycle of used fuel enables increased utilization of uranium resource and potential waste management benefits. - Once through fuel cycle uses less than 1% of energy value of the uranium. Courtesy AREVA 2 April 29, 2010 Summary of Fuel Cycle Options 3 Once-Through Fuel Cycle - One pass through reactor, used fuel directly disposed in a geologic repository. Modified Open Cycle - No or limited separations steps and

107

Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions  

SciTech Connect

Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm lithium metaborate solution respectively at the saturation temperature for 1000 psi (68.9 bar) coolant pressure. Boiling tests also revealed the formation of fine deposits of boron and lithium on the cladding surface which degraded the heat transfer rates. The boron and lithium metaborate precipitates after a 5 day test at 5000 ppm concentration and 1000 psi (68.9 bar) operating pressure reduced the heat transfer rate 21% and 30%, respectively for the two solutions.

Schultis, J., Kenneth; Fenton, Donald, L.

2006-10-20T23:59:59.000Z

108

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

109

Proposed replacement and operation of the anhydrous hydrogen fluoride supply and fluidized-bed reactor system at Building 9212. Draft environmental assessment  

SciTech Connect

The US Department of Energy (DOE) proposes to replace the existing anhydrous hydrogen fluoride (AHF) supply and fluidized-bed reactor systems for the Weapons Grade Highly Enriched Uranium Chemical Recovery and Recycle Facility, Building 9212, which is Iocated within the Y-12 Plant on DOE`s Oak Ridge Reservation in Oak Ridge, Tennessee. The current AHF supply and fluidized-bed reactor systems were designed and constructed more than 40 years ago. Because of their deteriorating condition, the corrosive nature of the materials processed, and the antiquated design philosophy upon which they are based, their long-term reliability cannot be assured. The current AHF supply system cannot mitigate an accidental release of AHF and vents fugitive AHF directly to the atmosphere during operations. the proposed action would reduce the risk of exposing the Y-12 Plant work force, the public, and the environment to an accidental release of AHF and would ensure the continuing ability of the Y-12 Plant to manufacture highly enriched uranium metal and process uranium from retired weapons for storage.

NONE

1995-03-01T23:59:59.000Z

110

DOE-NE Light Water Reactor Sustainability Program and EPRI Long-Term Operations Program – Joint Research and Development Plan  

SciTech Connect

Nuclear power has contributed almost 20% of the total amount of electricity generated in the United States over the past two decades. High capacity factors and low operating costs make nuclear power plants (NPPs) some of the most economical power generators available. Further, nuclear power remains the single largest contributor (nearly 70%) of non-greenhouse gas-emitting electric power generation in the United States. Even when major refurbishments are performed to extend operating life, these plants continue to represent cost-effective, low-carbon assets to the nation’s electrical generation capability.

Don Williams

2014-04-01T23:59:59.000Z

111

Computational evaluation of two reactor benchmark problems.  

E-Print Network (OSTI)

??A neutronic evaluation of two reactor benchmark problems was performed. The benchmark problems describe typical PWR uranium and plutonium (mixed oxide) fueled lattices. WIMSd4m, a… (more)

Cowan, James Anthony

2012-01-01T23:59:59.000Z

112

Light Water Reactor Sustainability Program Operator Performance Metrics for Control Room Modernization: A Practical Guide for Early Design Evaluation  

SciTech Connect

As control rooms are modernized with new digital systems at nuclear power plants, it is necessary to evaluate the operator performance using these systems as part of a verification and validation process. There are no standard, predefined metrics available for assessing what is satisfactory operator interaction with new systems, especially during the early design stages of a new system. This report identifies the process and metrics for evaluating human system interfaces as part of control room modernization. The report includes background information on design and evaluation, a thorough discussion of human performance measures, and a practical example of how the process and metrics have been used as part of a turbine control system upgrade during the formative stages of design. The process and metrics are geared toward generalizability to other applications and serve as a template for utilities undertaking their own control room modernization activities.

Ronald Boring; Roger Lew; Thomas Ulrich; Jeffrey Joe

2014-03-01T23:59:59.000Z

113

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

114

Extension of the supercritical carbon dioxide brayton cycle to low reactor power operation: investigations using the coupled anl plant dynamics code-SAS4A/SASSYS-1 liquid metal reactor code system.  

SciTech Connect

Significant progress has been made on the development of a control strategy for the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle enabling removal of power from an autonomous load following Sodium-Cooled Fast Reactor (SFR) down to decay heat levels such that the S-CO{sub 2} cycle can be used to cool the reactor until decay heat can be removed by the normal shutdown heat removal system or a passive decay heat removal system such as Direct Reactor Auxiliary Cooling System (DRACS) loops with DRACS in-vessel heat exchangers. This capability of the new control strategy eliminates the need for use of a separate shutdown heat removal system which might also use supercritical CO{sub 2}. It has been found that this capability can be achieved by introducing a new control mechanism involving shaft speed control for the common shaft joining the turbine and two compressors following reduction of the load demand from the electrical grid to zero. Following disconnection of the generator from the electrical grid, heat is removed from the intermediate sodium circuit through the sodium-to-CO{sub 2} heat exchanger, the turbine solely drives the two compressors, and heat is rejected from the cycle through the CO{sub 2}-to-water cooler. To investigate the effectiveness of shaft speed control, calculations are carried out using the coupled Plant Dynamics Code-SAS4A/SASSYS-1 code for a linear load reduction transient for a 1000 MWt metallic-fueled SFR with autonomous load following. No deliberate motion of control rods or adjustment of sodium pump speeds is assumed to take place. It is assumed that the S-CO{sub 2} turbomachinery shaft speed linearly decreases from 100 to 20% nominal following reduction of grid load to zero. The reactor power is calculated to autonomously decrease down to 3% nominal providing a lengthy window in time for the switchover to the normal shutdown heat removal system or for a passive decay heat removal system to become effective. However, the calculations reveal that the compressor conditions are calculated to approach surge such that the need for a surge control system for each compressor is identified. Thus, it is demonstrated that the S-CO{sub 2} cycle can operate in the initial decay heat removal mode even with autonomous reactor control. Because external power is not needed to drive the compressors, the results show that the S-CO{sub 2} cycle can be used for initial decay heat removal for a lengthy interval in time in the absence of any off-site electrical power. The turbine provides sufficient power to drive the compressors. Combined with autonomous reactor control, this represents a significant safety advantage of the S-CO{sub 2} cycle by maintaining removal of the reactor power until the core decay heat falls to levels well below those for which the passive decay heat removal system is designed. The new control strategy is an alternative to a split-shaft layout involving separate power and compressor turbines which had previously been identified as a promising approach enabling heat removal from a SFR at low power levels. The current results indicate that the split-shaft configuration does not provide any significant benefits for the S-CO{sub 2} cycle over the current single-shaft layout with shaft speed control. It has been demonstrated that when connected to the grid the single-shaft cycle can effectively follow the load over the entire range. No compressor speed variation is needed while power is delivered to the grid. When the system is disconnected from the grid, the shaft speed can be changed as effectively as it would be with the split-shaft arrangement. In the split-shaft configuration, zero generator power means disconnection of the power turbine, such that the resulting system will be almost identical to the single-shaft arrangement. Without this advantage of the split-shaft configuration, the economic benefits of the single-shaft arrangement, provided by just one turbine and lower losses at the design point, are more important to the overall cycle performance. Therefore, the single-shaft

Moisseytsev, A.; Sienicki, J. J. (Nuclear Engineering Division)

2012-05-10T23:59:59.000Z

115

Power Reactor Progress  

Science Journals Connector (OSTI)

Argonne kicks off EBWR; Allis-Chalmers plans power reactor using both nuclear and conventional fuels ... NUCLEAR POWER took two giant steps last week. ... Just as the first nuclear power system in the U. S. designed and built solely for the generation of electric power went into full operation at Argonne, Allis-Chalmers came up with a new twist in power reactors—a controlled recirculation boiling reactor (CRBR) using both nuclear and conventional fuels (C&EN, Feb. 18, page 7). ...

1957-02-25T23:59:59.000Z

116

Spectral Effects on Stress Relaxation of Inconel X-750 Springs in CANDU Reactors  

SciTech Connect

CANDU reactors have been operating for periods up to about 25 years. During this time there are changes to the nuclear reactor core components that are a function of operating environment and time. It is important to know how the properties of critical core components are likely to change over the life of a reactor and therefore their behaviours are characterised long before the end of the reactor design life. Tests are typically conducted in materials test reactors. The behaviour of a material is often characterised as a function of fast neutron fluence and the expected effect of operating time is established by simply extrapolating as a function of fluence. This may be appropriate when the neutron energy spectrum for the materials test reactor matches closely the neutron spectrum where the component resides in the power reactor. However, in cases where the spectrum is very different one has to convert the accumulated dose into a unit that is common in its effect on the material properties. For many property changes in nuclear reactor cores this unit is displacements per atom (dpa).

Griffiths, M.; Butcher, F. J.; Ariani, I.; Douglas, S.; Garner, Francis A.; Greenwood, Lawrence R.

2008-11-16T23:59:59.000Z

117

ORIGEN-ARP Cross-Section Libraries for Magnox, Advanced Gas-Cooled, and VVER Reactor Designs  

SciTech Connect

Cross-section libraries for the ORIGEN-ARP system were extended to include four non-U.S. reactor types: the Magnox reactor, the Advanced Gas-Cooled Reactor, the VVER-440, and the VVER-1000. Typical design and operational parameters for these four reactor types were determined by an examination of a variety of published information sources. Burnup simulation models of the reactors were then developed using the SAS2H sequence from the Oak Ridge National Laboratory SCALE code system. In turn, these models were used to prepare the burnup-dependent cross-section libraries suitable for use with ORIGEN-ARP. The reactor designs together with the development of the SAS2H models are described, and a small number of validation results using spent-fuel assay data are reported.

Murphy, BD

2004-03-10T23:59:59.000Z

118

Reactor Safety Planning for Prometheus Project, for Naval Reactors Information  

SciTech Connect

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

P. Delmolino

2005-05-06T23:59:59.000Z

119

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

120

Minimizing or eliminating refueling of nuclear reactor  

DOE Patents (OSTI)

Demand for refueling of a liquid metal fast nuclear reactor having a life of 30 years is eliminated or reduced to intervals of at least 10 years by operating the reactor at a low linear-power density, typically 2.5 kw/ft of fuel rod, rather than 7.5 or 15 kw/ft, which is the prior art practice. So that power of the same magnitude as for prior art reactors is produced, the volume of the core is increased. In addition, the height of the core and it diameter are dimensioned so that the ratio of the height to the diameter approximates 1 to the extent practicable considering the requirement of control and that the pressure drop in the coolant shall not be excessive. The surface area of a cylinder of given volume is a minimum if the ratio of the height to the diameter is 1. By minimizing the surface area, the leakage of neutrons is reduced. By reducing the linear-power density, increasing core volume, reducing fissile enrichment and optimizing core geometry, internal-core breeding of fissionable fuel is substantially enhanced. As a result, core operational life, limited by control worth requirements and fuel burnup capability, is extended up to 30 years of continuous power operation.

Doncals, Richard A. (Washington, PA); Paik, Nam-Chin (Pittsburgh, PA); Andre, Sandra V. (Hempfield Township, Westmoreland County, PA); Porter, Charles A. (Rostraver Township, Westmoreland County, PA); Rathbun, Roy W. (Greensburg, PA); Schwallie, Ambrose L. (Greensburg, PA); Petras, Diane S. (Penn Township, Westmoreland County, PA)

1989-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor operators typically" 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

Environmentally assisted cracking of light-water reactor materials  

SciTech Connect

Environmentally assisted cracking (EAC) of lightwater reactor (LWR) materials has affected nuclear reactors from the very introduction of the technology. Corrosion problems have afflicted steam generators from the very introduction of pressurized water reactor (PWR) technology. Shippingport, the first commercial PWR operated in the United States, developed leaking cracks in two Type 304 stainless steel (SS) steam generator tubes as early as 1957, after only 150 h of operation. Stress corrosion cracks were observed in the heat-affected zones of welds in austenitic SS piping and associated components in boiling-water reactors (BRWs) as early as 1965. The degradation of steam generator tubing in PWRs and the stress corrosion cracking (SCC) of austenitic SS piping in BWRs have been the most visible and most expensive examples of EAC in LWRs, and the repair and replacement of steam generators and recirculation piping has cost hundreds of millions of dollars. However, other problems associated with the effects of the environment on reactor structures and components am important concerns in operating plants and for extended reactor lifetimes. Cast duplex austenitic-ferritic SSs are used extensively in the nuclear industry to fabricate pump casings and valve bodies for LWRs and primary coolant piping in many PWRs. Embrittlement of the ferrite phase in cast duplex SS may occur after 10 to 20 years at reactor operating temperatures, which could influence the mechanical response and integrity of pressure boundary components during high strain-rate loading (e.g., seismic events). The problem is of most concern in PWRs where slightly higher temperatures are typical and cast SS piping is widely used.

Chopra, O.K.; Chung, H.M.; Kassner, T.F.; Shack, W.J.

1996-02-01T23:59:59.000Z

122

A pilot reactor study to determine operational factors of the commercial hydrodesulphurization (HDS) catalyst to produce ultra-low sulphur diesel (ULSD)  

Science Journals Connector (OSTI)

Abstract A pilot plant test was carried out using a catalyst from the refinery hydrotreater unit to process blend feed of 70 vol% Light Gas Oil – LGO & 30 vol% hydrotreated Gas Oil – HDT GO to produce 8 ppm ultra-low sulphur diesel (ULSD) product. The impact of changes in process conditions have been studied for catalyst deactivation rate and hydrogen consumption. Test results shown that refinery unit can process such blend feed to reach 8 ppm sulphur product by an increase of the reactor bed temperature. However due to higher aromatic content in the blend feed as compare to reference feed LGO, moderate increase in hydrogen consumption was also observed. Catalyst performance was evaluated at 55/42/32 bar hydrogen partial pressure (PPH2) to determine catalyst deactivation rate and hydrogen consumption, targeting ULSD product. A decrease of PPH2 from 52 bar (which is current operating condition) to 32 bar resulted in reduction of H2 consumption but also shows decrease of catalyst cycle length due to higher deactivation rate. Pilot plant test shows that by contriving computational methods and analysis techniques for hydrogen balance & catalyst deactivation rate from the pilot plant test data, it becomes possible to predict catalyst performance in commercial unit.

Nilesh Chandak; Adel Al Hamadi; Mohamed Yousef; Abdulhamid Mohamed; Kazuhiro Inamura; Mikael Berthod

2014-01-01T23:59:59.000Z

123

Improve reformer operation with trace sulfur removal  

SciTech Connect

Modern bimetallic reforming catalysts typically have feed specifications for sulfur of 0.5 to 1 wppm in the reformer naphtha carge. Sulfur in the raw naphtha is reduced to this level by naphtha hydrotreating. While most naphtha hydrotreating operations can usually obtain these levels without substantial problems. It is difficult to obtain levels much below 0.5 to 1 wppm with this process. Revamp of a constrained existing hydrotreater to reduce product sulfur slightly can be extremely costly typically entailing replacement or addition of a new reactor. At Engelhard the authors demonstrated that if the last traces of sulfur remaining from hydrotreating can be removed, the resulting ultra-low sulfur feed greatly improves the reformer operation and provides substantial economic benefit to the refiner. Removal of the remaining trace sulfur is accomplished in a simple manner with a special adsorbent bed, without adding complexity to the reforming operation.

McClung, R.G.; Novak, W.J.

1987-01-01T23:59:59.000Z

124

INL @ work: Nuclear Reactor Operator  

ScienceCinema (OSTI)

INL @ work features jobs at the Idaho National Laboratory. Learn more about careers and energy research at INL's facebook site http://www.facebook.com/idahonationallaboratory

Russell, Patty

2013-05-28T23:59:59.000Z

125

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

126

International Journal of Chemical Reactor Engineering  

E-Print Network (OSTI)

International Journal of Chemical Reactor Engineering Volume 3 2005 Article A17 Optimal Operation, a single re- action takes place in the reactor and the operational objective is to compute the optimal feed is illustrated via simulation of two semi-batch reactor applications. KEYWORDS: Dynamic Optimization, Batch

Palanki, Srinivas

127

Canadian university research reactors  

SciTech Connect

In Canada there are seven university research reactors: one medium-power (2-MW) swimming pool reactor at McMaster University and six low-power (20-kW) SLOWPOKE reactors at Dalhousie University, Ecole Polytechnique, the Royal Military College, the University of Toronto, the University of Saskatchewan, and the University of Alberta. This paper describes primarily the McMaster Nuclear Reactor (MNR), which operates on a wider scale than the SLOWPOKE reactors. The MNR has over a hundred user groups and is a very broad-based tool. The main applications are in the following areas: (1) neutron activation analysis (NAA); (2) isotope production; (3) neutron beam research; (4) nuclear engineering; (5) neutron radiography; and (6) nuclear physics.

Ernst, P.C.; Collins, M.F.

1989-11-01T23:59:59.000Z

128

Unique features of space reactors  

SciTech Connect

Space reactors are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K. 8 refs., 3 figs., 1 tab.

Buden, D.

1990-01-01T23:59:59.000Z

129

Alternate-fuel reactor studies  

SciTech Connect

A number of studies related to improvements and/or greater understanding of alternate-fueled reactors is presented. These studies cover the areas of non-Maxwellian distributions, materials and lifetime analysis, a /sup 3/He-breeding blanket, tritium-rich startup effects, high field magnet support, and reactor operation spanning the range from full D-T operation to operation with no tritium breeding.

Evans, K. Jr.; Ehst, D.A.; Gohar, Y.; Jung, J.; Mattas, R.F.; Turner, L.R.

1983-02-01T23:59:59.000Z

130

INTEGRATION OF HIGH TEMPERATURE GAS REACTORS WITH IN SITU OIL SHALE RETORTING  

SciTech Connect

This paper evaluates the integration of a high-temperature gas-cooled reactor (HTGR) to an in situ oil shale retort operation producing 7950 m3/D (50,000 bbl/day). The large amount of heat required to pyrolyze the oil shale and produce oil would typically be provided by combustion of fossil fuels, but can also be delivered by an HTGR. Two cases were considered: a base case which includes no nuclear integration, and an HTGR-integrated case.

Eric P. Robertson; Michael G. McKellar; Lee O. Nelson

2011-05-01T23:59:59.000Z

131

Numerical study of hydrogen production by the sorption-enhanced steam methane reforming process with online CO2 capture as operated in fluidized bed reactors  

Science Journals Connector (OSTI)

A three-dimensional (3D) Eulerian two-fluid model with an in-house code was developed to simulate the gas-particle two-phase flow in the fluidized bed reactors. The CO2 capture with Ca-based sorbents in the steam

Yuefa Wang; Zhongxi Chao; Hugo A. Jakobsen

2011-08-01T23:59:59.000Z

132

Reactor Safety Research Programs  

SciTech Connect

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

Edler, S. K.

1981-07-01T23:59:59.000Z

133

Dust Divertor for a Tokamak Fusion Reactor  

Science Journals Connector (OSTI)

The conventional tokamak fusion reactor deploys a magnetic divertor design which channels...1], or covered by flowing liquid metals [2...]. A typical estimate for the plasma heat flux to the divertor for a tokama...

X. Z. Tang; G. L. Delzanno

2010-10-01T23:59:59.000Z

134

Thermonuclear Reflect AB-Reactor  

E-Print Network (OSTI)

The author offers a new kind of thermonuclear reflect reactor. The remarkable feature of this new reactor is a three net AB reflector, which confines the high temperature plasma. The plasma loses part of its energy when it contacts with the net but this loss can be compensated by an additional permanent plasma heating. When the plasma is rarefied (has a small density), the heat flow to the AB reflector is not large and the temperature in the triple reflector net is lower than 2000 - 3000 K. This offered AB-reactor has significantly less power then the currently contemplated power reactors with magnetic or inertial confinement (hundreds-thousands of kW, not millions of kW). But it is enough for many vehicles and ships and particularly valuable for tunnelers, subs and space apparatus, where air to burn chemical fuel is at a premium or simply not available. The author has made a number of innovations in this reactor, researched its theory, developed methods of computation, made a sample computation of typical project. The main point of preference for the offered reactor is its likely cheapness as a power source. Key words: Micro-thermonuclear reactor, Multi-reflex AB-thermonuclear reactor, Self-magnetic AB-thermonuclear reactor, aerospace thermonuclear engine.

Alexander Bolonkin

2008-03-26T23:59:59.000Z

135

Ignition reactor and pump pulse parameters in a reactor–laser system  

Science Journals Connector (OSTI)

The experience gained in operating a demonstration nuclear-pumped laser in stand B (Physics and Power- Engineering Institute (FEI)) with a pulsed ignition reactor based on the 235U BARS-6 reactor is analyzed. It ...

P. P. D’yachenko; G. N. Fokin

2012-09-01T23:59:59.000Z

136

Manipulating decision making of typical agents  

E-Print Network (OSTI)

We investigate how the choice of decision makers can be varied under the presence of risk and uncertainty. Our analysis is based on the approach we have previously applied to individual decision makers, which we now generalize to the case of decision makers that are members of a society. The approach employs the mathematical techniques that are common in quantum theory, justifying our naming as Quantum Decision Theory. However, we do not assume that decision makers are quantum objects. The techniques of quantum theory are needed only for defining the prospect probabilities taking into account such hidden variables as behavioral biases and other subconscious feelings. The approach describes an agent's choice as a probabilistic event occurring with a probability that is the sum of a utility factor and of an attraction factor. The attraction factor embodies subjective and unconscious dimensions in the mind of the decision maker. We show that the typical aggregate amplitude of the attraction factor is $1/4$, and ...

Yukalov, V I

2014-01-01T23:59:59.000Z

137

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

138

N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor Cocooning  

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

N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor Cocooning Project Now Complete N Reactor Placed In Interim Safe Storage: Largest Hanford Reactor Cocooning Project Now Complete June 14, 2012 - 12:00pm Addthis Media Contacts Cameron Hardy Cameron.Hardy@rl.doe.gov 509-376-5365 Mark McKenna mmckenna@wch-rcc.com 509-372-9032 RICHLAND, WASH. - The U.S. Department of Energy's (DOE's) River Corridor contractor, Washington Closure Hanford, has completed placing N Reactor in interim safe storage, a process also known as "cocooning." N Reactor was the last of nine plutonium production reactors to be shut down at DOE's Hanford Site in southeastern Washington state. It was Hanford's longest-running reactor, operating from 1963 to 1987. "In the 1960's, N Reactor represented the future of energy in America.

139

Nuclear reactor multiphysics via bond graph formalism  

E-Print Network (OSTI)

This work proposes a simple and effective approach to modeling nuclear reactor multiphysics problems using bond graphs. Conventional multiphysics simulation paradigms normally use operator splitting, which treats the ...

Sosnovsky, Eugeny

2014-01-01T23:59:59.000Z

140

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

Note: This page contains sample records for the topic "reactor operators typically" 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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Assessment of High Value Surveillance Materials Assessment of High Value Surveillance Materials Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials 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 that 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 pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely

142

Temperature effects on chemical reactor  

Science Journals Connector (OSTI)

In this paper we had to study some characteristics of the chemical reactors from which we can understand the reactor operation in different circumstances; from these and the most important factor that has a great effect on the reactor operation is the temperature it is a mathematical processing of a chemical problem that was already studied but it may be developed by introducing new strategies of control; in our case we deal with the analysis of a liquid?gas reactor which can make the flotation of the benzene to produce the ethylene; this type of reactors can be used in vast domains of the chemical industry especially in refinery plants where we find the oil separation and its extractions whether they are gases or liquids which become necessary for industrial technology especially in our century.

M. Azzouzi

2008-01-01T23:59:59.000Z

143

NUCLEAR REACTORS.  

E-Print Network (OSTI)

??Nuclear reactors are devices containing fissionable material in sufficient quantity and so arranged as to be capable of maintaining a controlled, self-sustaining NUCLEAR FISSION chain… (more)

Belachew, Dessalegn

2010-01-01T23:59:59.000Z

144

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

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

Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Conduct of Operations - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR February...

145

Manipulating decision making of typical agents  

E-Print Network (OSTI)

We investigate how the choice of decision makers can be varied under the presence of risk and uncertainty. Our analysis is based on the approach we have previously applied to individual decision makers, which we now generalize to the case of decision makers that are members of a society. The approach employs the mathematical techniques that are common in quantum theory, justifying our naming as Quantum Decision Theory. However, we do not assume that decision makers are quantum objects. The techniques of quantum theory are needed only for defining the prospect probabilities taking into account such hidden variables as behavioral biases and other subconscious feelings. The approach describes an agent's choice as a probabilistic event occurring with a probability that is the sum of a utility factor and of an attraction factor. The attraction factor embodies subjective and unconscious dimensions in the mind of the decision maker. We show that the typical aggregate amplitude of the attraction factor is $1/4$, and it can be either positive or negative depending on the relative attraction of the competing choices. The most efficient way of varying the decision makers choice is realized by influencing the attraction factor. This can be done in two ways. One method is to arrange in a special manner the payoff weights, which induces the required changes of the values of attraction factors. We show that a slight variation of the payoff weights can invert the sign of the attraction factors and reverse the decision preferences, even when the prospect utilities remain unchanged. The second method of influencing the decision makers choice is by providing information to decision makers. The methods of influencing decision making are illustrated by several experiments, whose outcomes are compared quantitatively with the predictions of our approach.

V. I. Yukalov; D. Sornette

2014-09-02T23:59:59.000Z

146

Control of reactor coolant flow path during reactor decay heat removal  

DOE Patents (OSTI)

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Hunsbedt, Anstein N. (Los Gatos, CA)

1988-01-01T23:59:59.000Z

147

Analysis of the operational reliability of a power-generating unit with a BN-600 reactor during the period 1980–1993  

Science Journals Connector (OSTI)

The high quality of the design and the additional improvements to separate units of the main equipment and systems at the initial stage of operation (first main circulation pump, steam generators, safety and c...

N. N. Oshkanov; A. G. Sheinkman; P. P. Govorov

1994-03-01T23:59:59.000Z

148

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

149

Knowledge base expert system control of spatial xenon oscillations in pressurized water reactors  

Science Journals Connector (OSTI)

Current xenon oscillation control methods used in pressurized water reactors are knowledge intensive, and heuristic in nature. An expert system is developed to implement the heuristic constant axial offset control procedure to aid reactor operators in increasing the plant reliability by reducing the human error component of the failure probability. An expert system is written in the production system language, OPS5, with a forward chaining algorithm. It samples the reactor core with a certain time interval, evaluates the core status to determine the necessary corrective actions in terms of reactivity insertion, and selects the control parameter to realize this reactivity insertion. The amount of control action is determined using a knowledge base which consists of the differential rod worth curves, and the boron reactivity worth of a given reactor. The controller has been tested using a one-dimesional core model for verification of the rules and the code. It has been shown that, having the reactor dependent parameters in its knowledge base, the controller is able to follow a typical load demand for a daily cycle of a reactor, and is able to keep the axial offset within a target band.

Serhat Alten; Richard A Danofsky

1993-01-01T23:59:59.000Z

150

History of Research Reactors at Brookhaven  

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

History of Research Reactors at Brookhaven History of Research Reactors at Brookhaven Brookhaven National Laboratory has three nuclear reactors on its site that were used for scientific research. The reactors are all shut down, and the Laboratory is addressing environmental issues associated with their operations. photo of BGRR Brookhaven Graphite Research Reactor - Beginning operations in 1950, the graphite reactor was used for research in medicine, biology, chemistry, physics and nuclear engineering. One of the most significant achievements at this facility was the development of technetium-99m, a radiopharmaceutical widely used to image almost any organ in the body. The graphite reactor was shut down in 1969. Parts of it have been decommissioned, with the remainder to be addressed by 2011. More history

151

Nuclear power reactor education and training at the Ford nuclear reactor  

SciTech Connect

Since 1977, staff members of the University of Michigan's Ford nuclear reactor have provided courses and reactor laboratory training programs for reactor operators, engineers, and technicians from seven electric utilities, including Cleveland Electric Illuminating, Consumers Power, Detroit Edison, Indiana and Michigan Electric, Nebraska Public Power, Texas Utilities Generating Company, and Toledo Edison. Reactor laboratories, instrument technician training, and reactor physics courses have been conducted at the university. Courses conducted at plant sites include reactor physics, thermal sciences, materials sciences, and health physics and radiation protection.

Burn, R.R.

1989-01-01T23:59:59.000Z

152

Light Water Reactor Sustainability Technical Documents | Department of  

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

Reactor Technologies » Light Water Reactor Reactor Technologies » Light Water Reactor Sustainability Program » Light Water Reactor Sustainability Technical Documents Light Water Reactor Sustainability Technical Documents April 30, 2013 LWRS Program and EPRI Long-Term Operations Program - Joint R&D Plan To address the challenges associated with pursuing commercial nuclear power plant operations beyond 60 years, the U.S. Department of Energy's (DOE) Office of Nuclear Energy (NE) and the Electric Power Research Institute (EPRI) have established separate but complementary research and development programs: DOE-NE's Light Water Reactor Sustainability (LWRS) Program and EPRI's Long-Term Operations (LTO) Program. April 30, 2013 Light Water Reactor Sustainability Program - Integrated Program Plan The Light Water Reactor Sustainability (LWRS) Program is a research and

153

Refueling Simulation Strategy of a CANDU Reactor Based on Optimum Zone Controller Water Levels  

SciTech Connect

An optimum refueling simulation method was developed for application to a Canada deuterium uranium 713-MW(electric) (CANDU-6) reactor. The objective of the optimization was to maintain the operating range of the zone controller unit (ZCU) water level so that the reference zone power distribution is reproduced following the refueling operation. The zone controller level on the refueling operation was estimated by the generalized perturbation method, which provides sensitivities of the zone power to an individual refueling operation and the zone controller level. By constructing a system equation of the zone power, the zone controller level was obtained, which was used to find the most suitable combination of the refueling channels. The 250-full-power-day refueling simulations showed that the channel and bundle powers are well controlled below the license limits when the ZCU water level remains in the typical operating range.

Choi, Hangbok; Kim, Do Heon [Korea Atomic Energy Research Institute (Korea, Republic of)

2005-09-15T23:59:59.000Z

154

Design of chemical reactors of the heat exchanger type  

E-Print Network (OSTI)

Operating Profile - Example I 23 , 53 Heat Rate Comparison - Example I Operating Profile - Example 2 Operating Profile - Example 3 Operating Profile - Example 4 Equations (113) and (114) at 790 Reactor Profile - Exan piss 5 and 6 Heat of Reaction.... simple inathematical function of time. While his work was a step forward, it is not directly applicable to the problem of reactor design. Hougen and Watsor. (3), and recently Fair and Rase (4), illustra- ted an exact non-machine method of reactor...

McBeth, Lloyd Theodore

2012-06-07T23:59:59.000Z

155

Human Reliability Considerations for Small Modular Reactors  

SciTech Connect

Small modular reactors (SMRs) are a promising approach to meeting future energy needs. Although the electrical output of an individual SMR is relatively small compared to that of typical commercial nuclear plants, they can be grouped to produce as much energy as a utility demands. Furthermore, SMRs can be used for other purposes, such as producing hydrogen and generating process heat. The design characteristics of many SMRs differ from those of current conventional plants and may require a distinct concept of operations. The U.S. Nuclear Regulatory Commission (NRC) conducted research to examine the human factors engineering and the operational aspects of SMRs. The research identified thirty potential human-performance issues that should be considered in the NRC's reviews of SMR designs and in future research activities. The purpose of this report is to illustrate how the issues can support SMR probabilistic risk analyses and their review by identifying potential human failure events for a subset of the issues. As part of addressing the human contribution to plant risk, human reliability analysis practitioners identify and quantify the human failure events that can negatively impact normal or emergency plant operations. The results illustrated here can be generalized to identify additional human failure events for the issues discussed and can be applied to those issues not discussed in this report.

OHara J. M.; Higgins, H.; DAgostino, A.; Erasmia, L.

2012-01-27T23:59:59.000Z

156

TABLE 2. U.S. Nuclear Reactor Ownership Data  

U.S. Energy Information Administration (EIA) Indexed Site

2. U.S. Nuclear Reactor Ownership Data" "PlantReactor Name","Generator ID","Utility Name - Operator","Owner Name","% Owned" "Arkansas Nuclear One",1,"Entergy Arkansas...

157

Experimental and Computational Study of Fluid Dynamics in Solar Reactor  

E-Print Network (OSTI)

The experimental simulation and a computational validation of a methane-cracking solar reactor powered by solar energy is the focus of this article. A solar cyclone reactor operates at over 1000 °C where the methane decomposition reaction takes...

Chien, Min-Hsiu

2014-02-19T23:59:59.000Z

158

A comparison of nuclear reactor control room display panels  

E-Print Network (OSTI)

complex and time consuming task. It is expected that the control room of future commercial nuclear reactor power plants will change considerably as a result of these studies. Currently there are literally hundreds of displays and controls...: Dr. Rodger S. Koppa A study was conducted to investigate the use of computer generated displays to operate nuclear reactor power plants. The AGN-201 reactor at Texas A&M university was the reactor studied. After observing several licensed reactor...

Bowers, Frances Renae

2012-06-07T23:59:59.000Z

159

Light Water Reactor Sustainability  

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

4 Light Water Reactor Sustainability ACCOMPLISHMENTS REPORT 2014 Accomplishments Report | Light Water Reactor Sustainability 2 T he mission of the Light Water Reactor...

160

Reactor antineutrino monitoring with a plastic scintillator array as a new safeguards method  

E-Print Network (OSTI)

We developed a segmented reactor-antineutrino detector made of plastic scintillators for application as a tool in nuclear safeguards inspection and performed mostly unmanned field operations at a commercial power plant reactor. At a position outside the reactor building, we measured the difference in reactor antineutrino flux above the ground when the reactor was active and inactive.

S. Oguri; Y. Kuroda; Y. Kato; R. Nakata; Y. Inoue; C. Ito; M. Minowa

2014-05-23T23:59:59.000Z

Note: This page contains sample records for the topic "reactor operators typically" 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

Reactor antineutrino monitoring with a plastic scintillator array as a new safeguards method  

Science Journals Connector (OSTI)

Abstract We developed a segmented reactor-antineutrino detector made of plastic scintillators for application as a tool in nuclear safeguards inspection and performed mostly unmanned field operations at a commercial power plant reactor. At a position outside the reactor building, we measured the difference in reactor antineutrino flux above the ground when the reactor was active and inactive.

S. Oguri; Y. Kuroda; Y. Kato; R. Nakata; Y. Inoue; C. Ito; M. Minowa

2014-01-01T23:59:59.000Z

162

Accelerators for Subcritical Molten-Salt Reactors  

SciTech Connect

Accelerator parameters for subcritical reactors have usually been based on using solid nuclear fuel much like that used in all operating critical reactors as well as the thorium burning accelerator-driven energy amplifier proposed by Rubbia et al. An attractive alternative reactor design that used molten salt fuel was experimentally studied at ORNL in the 1960s, where a critical molten salt reactor was successfully operated using enriched U235 or U233 tetrafluoride fuels. These experiments give confidence that an accelerator-driven subcritical molten salt reactor will work better than conventional reactors, having better efficiency due to their higher operating temperature, having the inherent safety of subcritical operation, and having constant purging of volatile radioactive elements to eliminate their accumulation and potential accidental release in dangerous amounts. Moreover, the requirements to drive a molten salt reactor can be considerably relaxed compared to a solid fuel reactor, especially regarding accelerator reliability and spallation neutron targetry, to the point that much of the required technology exists today. It is proposed that Project-X be developed into a prototype commercial machine to produce energy for the world by, for example, burning thorium in India and nuclear waste from conventional reactors in the USA.

Johnson, Roland (Muons, Inc.) [Muons, Inc.

2011-08-03T23:59:59.000Z

163

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

164

Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency  

SciTech Connect

Suggested for Track 7: Advances in Reactor Core Design and In-Core Management _____________________________________________________________________________________ Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency R. Wigeland and K. Hamman Idaho National Laboratory Given the ability of fast reactors to effectively transmute the transuranic elements as are present in spent nuclear fuel, fast reactors are being considered as one element of future nuclear power systems to enable continued use and growth of nuclear power by limiting high-level waste generation. However, a key issue for fast reactors is higher electricity cost relative to other forms of nuclear energy generation. The economics of the fast reactor are affected by the amount of electric power that can be produced from a reactor, i.e., the thermal efficiency for electricity generation. The present study is examining the potential for fast reactor subassembly design changes to improve the thermal efficiency by increasing the average coolant outlet temperature without increasing peak temperatures within the subassembly, i.e., to make better use of current technology. Sodium-cooled fast reactors operate at temperatures far below the coolant boiling point, so that the maximum coolant outlet temperature is limited by the acceptable peak temperatures for the reactor fuel and cladding. Fast reactor fuel subassemblies have historically been constructed using a large number of small diameter fuel pins contained within a tube of hexagonal cross-section, or hexcan. Due to this design, there is a larger coolant flow area next to the hexcan wall as compared to flow area in the interior of the subassembly. This results in a higher flow rate near the hexcan wall, overcooling the fuel pins next to the wall, and a non-uniform coolant temperature distribution. It has been recognized for many years that this difference in sodium coolant temperature was detrimental to achieving greater thermal efficiency, since it causes the fuel pins in the center of the subassembly to operate at higher temperatures than those near the hexcan walls, and it is the temperature limit(s) for those fuel pins that limits the average coolant outlet temperature. Fuel subassembly design changes are being investigated using computational fluid dynamics (CFD) to quantify the effect that the design changes have on reducing the intra-subassembly coolant flow and temperature distribution. Simulations have been performed for a 19-pin test subassembly geometry using typical fuel pin diameters and wire wrap spacers. The results have shown that it may be possible to increase the average coolant outlet temperature by 20 C or more without changing the peak temperatures within the subassembly. These design changes should also be effective for reactor designs using subassemblies with larger numbers of fuel pins. R. Wigeland, Idaho National Laboratory, P.O. Box 1625, Mail Stop 3860, Idaho Falls, ID, U.S.A., 83415-3860 email – roald.wigeland@inl.gov fax (U.S.) – 208-526-2930

R. Wigeland; K. Hamman

2009-09-01T23:59:59.000Z

165

An Overview of the Safety Case for Small Modular Reactors  

SciTech Connect

Several small modular reactor (SMR) designs emerged in the late 1970s and early 1980s in response to lessons learned from the many technical and operational challenges of the large Generation II light-water reactors. After the accident at the Three Mile Island plant in 1979, an ensuing reactor redesign effort spawned the term inherently safe designs, which later evolved into passively safe terminology. Several new designs were engineered to be deliberately small in order to fully exploit the benefits of passive safety. Today, new SMR designs are emerging with a similar philosophy of offering highly robust and resilient designs with increased safety margins. Additionally, because these contemporary designs are being developed subsequent to the September 11, 2001, terrorist attack, they incorporate a number of intrinsic design features to further strengthen their safety and security. Several SMR designs are being developed in the United States spanning the full spectrum of reactor technologies, including water-, gas-, and liquid-metal-cooled ones. Despite a number of design differences, most of these designs share a common set of design principles to enhance plant safety and robustness, such as eliminating plant design vulnerabilities where possible, reducing accident probabilities, and mitigating accident consequences. An important consequence of the added resilience provided by these design approaches is that the individual reactor units and the entire plant should be able to survive a broader range of extreme conditions. This will enable them to not only ensure the safety of the general public but also help protect the investment of the owner and continued availability of the power-generating asset. Examples of typical SMR design features and their implications for improved plant safety are given for specific SMR designs being developed in the United States.

Ingersoll, Daniel T [ORNL] [ORNL

2011-01-01T23:59:59.000Z

166

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Initial Assessment of Thermal Annealing Needs and Challenges 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 40y to 80y implies a doubling of the neutron exposure for the RPV. Thus,

167

Nuclear reactor fissile isotopes antineutrino spectra  

E-Print Network (OSTI)

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

V. Sinev

2012-07-30T23:59:59.000Z

168

The Gas Reactor Makes a Comeback  

Science Journals Connector (OSTI)

...while the operators of a gas reactor can leave the...ofhis high 699 temperature gas-cooled reactor (HTGR...from the highly pressured turbine drive system might get...would produce combustible gas-es, creating the potential...too much to complete the remaining contracts. So General...

ELIOT MARSHALL

1984-05-18T23:59:59.000Z

169

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

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

has designed and operated 52 test reactors, including EBR-1, the world's first nuclear power plant Key Contributions System safety analysis Multiscale fuel performance...

170

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

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

the better understanding of the system uncertainties and sensitivities afforded by the virtual reactor will identify improvements in both the operation and design of the fuel...

171

Light Water Reactor Sustainability Newsletter Rebecca Smith-Kevern  

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

Rebecca Smith-Kevern Director, Office of Light Water Reactor Technologies. I am often asked why the Federal Government should fund a program that supports the continued operation...

172

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

173

Passive heat transfer means for nuclear reactors  

DOE Patents (OSTI)

An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. Means such as shrouding normally isolated the secondary condensing section from effective heat transfer with the heat sink, but a sensor responds to overheat conditions of the reactor to open the shrouding, which thereby increases the cooling capacity of the heat pipe. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

Burelbach, James P. (Glen Ellyn, IL)

1984-01-01T23:59:59.000Z

174

Rethinking the light water reactor fuel cycle  

E-Print Network (OSTI)

The once through nuclear fuel cycle adopted by the majority of countries with operating commercial power reactors imposes a number of concerns. The radioactive waste created in the once through nuclear fuel cycle has to ...

Shwageraus, Evgeni, 1973-

2004-01-01T23:59:59.000Z

175

Tritium production analysis and management strategies for a Fluoride-salt-cooled high-temperature test reactor (FHTR)  

E-Print Network (OSTI)

The Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a test reactor concept that aims to demonstrate the neutronics, thermal-hydraulics, materials, tritium management, and to address other reactor operational ...

Rodriguez, Judy N

2013-01-01T23:59:59.000Z

176

Radiation Damage and Tritium Breeding Study in a Fusion Reactor Using a Liquid Wall of Various Thorium Molten Salts  

Science Journals Connector (OSTI)

A new magnetic fusion reactor design, called APEX uses a liquid wall between fusion plasma and solid first wall to reach ... replacement of the first wall structure during the reactor’s operation due to the radia...

Mustafa Übeyli

2007-12-01T23:59:59.000Z

177

Typical Oak Ridge cemesto houses and city bus | Y-12 National...  

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

Typical Oak Ridge cemesto ... Typical Oak Ridge cemesto houses and city bus Typical Oak Ridge cemesto houses and city bus...

178

Computer aided nuclear reactor modeling  

E-Print Network (OSTI)

Nuclear reactor modeling is an important activity that lets us analyze existing as well as proposed systems for safety, correct operation, etc. The quality of a analysis is directly proportional to the quality of the model used. In this work we look...

Warraich, Khalid Sarwar

2012-06-07T23:59:59.000Z

179

Inherently Safe Reactors and a Second Nuclear Era  

Science Journals Connector (OSTI)

...Institute of Nuclear Power Operations. Improvements of this...reactors appear to be as safe as large dams, most...that, yes, inherently safe re-actors are surprisingly...more coal under utility boilers. But the environmental...parks. Would inherently safe reactors render existing...

Alvin M. Weinberg; Irving Spiewak

1984-06-29T23:59:59.000Z

180

Trickle - Bed reactor simulation using a process simulator  

Science Journals Connector (OSTI)

The present study deals with a multiple reaction system in both gas and liquid phases considering the effect of gas-liquid mass transfer limitations in a trickle-bed reactor where the catalytic hydrotreating of gas oil reaction is being carried out. ... Keywords: Trickle - bed reactor, hydrogenation, reactor model, user-added unit operation

E. Verruschi; J. Freitez; Y. Gonzalez; C. G. Dassori

2009-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor operators typically" from the National Library of EnergyBeta (NLEBeta).
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181

Overview of Sandia National Laboratories pulse nuclear reactors  

SciTech Connect

Sandia National Laboratories has designed, constructed and operated bare metal Godiva-type and pool-type pulse reactors since 1961. The reactor facilities were designed to support a wide spectrum of research, development, and testing activities associated with weapon and reactor systems.

Schmidt, T.R. [Sandia National Labs., Albuquerque, NM (United States); Reuscher, J.A. [Texas A& M Univ., College Station, TX (United States)

1994-10-01T23:59:59.000Z

182

Nuclear reactor control apparatus  

DOE Patents (OSTI)

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

Sridhar, Bettadapur N. (Cupertino, CA)

1983-11-01T23:59:59.000Z

183

Development of a system model for advanced small modular reactors.  

SciTech Connect

This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandia's concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

2014-01-01T23:59:59.000Z

184

X-10 Graphite Reactor | Department of Energy  

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

X-10 Graphite Reactor X-10 Graphite Reactor X-10 Graphite Reactor X-10 Graphite Reactor When President Roosevelt in December 1942 authorized the Manhattan Project, the Oak Ridge site in eastern Tennessee had already been obtained and plans laid for an air-cooled experimental pile, a pilot chemical separation plant, and support facilities. The X-10 Graphite Reactor, designed and built in ten months, went into operation on November 4, 1943. The X-10 used neutrons emitted in the fission of uranium-235 to convert uranium-238 into a new element, plutonium-239. The reactor consists of a huge block of graphite, measuring 24 feet on each side, surrounded by several feet of high-density concrete as a radiation shield. The block is pierced by 1,248 horizontal diamond-shaped channels in

185

Ordered bed modular reactor design proposal  

SciTech Connect

The Ordered Bed Modular Reactor (OBMR) is a design as an advanced modular HTGR in which the annular reactor core is filled with an ordered bed of fuel spheres. This arrangement allows fuel elements to be poured into the core cavity which is shaped so that an ordered bed is formed and to be discharged from the core through the opening holes in the reactor top. These operations can be performed in a shutdown shorter time. The OBMR has the most of advantages from both the pebble bed reactor and block type reactor. Its core has great structural flexibility and stability, which allow increasing reactor output power and outlet gas temperature as well as decreasing core pressure drop. This paper introduces ordered packing bed characteristics, unloading and loading technique of the fuel spheres and predicted design features of the OBMR. (authors)

Tian, J. [Inst. of Nuclear Energy Technology, Tsinghua Univ., Beijing 100084 (China)

2006-07-01T23:59:59.000Z

186

Evaluation of Torsatrons as reactors  

SciTech Connect

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors. This scoping study, which uses an integrated cost-minimization code that incorporates costing and reactor component models self-consistently with a 1-D energy transport calculation, shows that a torsatron reactor could also be economically competitive with a tokamak reactor. The projected cost of electricity (COE) estimated using the Advanced Reactor Innovation and Evaluation Studies (ARIES) costing algorithms is 65.6 mill/kW(e)h in constant 1992 dollars for a reference 1-GW(e) Compact Torsatron reactor case. The COE is relatively insensitive (<10% variation) over a wide range of assumptions, including variations in the maximum field allowed on the coils, the coil elongation, the shape of the density profile, the beta limit, the confinement multiplier, and the presence of a large loss region for alpha particles. The largest variations in the COE occur for variations in the electrical power output demanded and the plasma-coil separation ratio.

Lyon, J.F. [Oak Ridge National Lab., TN (United States); Gulec, K. [Univ. of Tennessee, Knoxville, TN (United States); Miller, R.L. [Los Alamos National Lab., NM (United States); El-Guebaly, L. [Univ. of Wisconsin, Madison, WI (United States)

1994-03-01T23:59:59.000Z

187

Rapid starting methanol reactor system  

DOE Patents (OSTI)

The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

Chludzinski, Paul J. (38 Berkshire St., Swampscott, MA 01907); Dantowitz, Philip (39 Nancy Ave., Peabody, MA 01960); McElroy, James F. (12 Old Cart Rd., Hamilton, MA 01936)

1984-01-01T23:59:59.000Z

188

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

189

Photocatalytic reactor  

DOE Patents (OSTI)

A photocatalytic reactor is described for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane. 4 figs.

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

1999-01-19T23:59:59.000Z

190

Signatures of Heating and Cooling Energy Consumption for Typical AHUs  

E-Print Network (OSTI)

An analysis is performed to investigate the signatures of different parameters on the heating and cooling energy consumption of typical air handling units (AHUs). The results are presented in graphic format. HVAC simulation engineers can use...

Wei, G.; Liu, M.; Claridge, D. E.

1998-01-01T23:59:59.000Z

191

Fusion reactor systems  

Science Journals Connector (OSTI)

In this review we consider deuterium-tritium (D-T) fusion reactors based on four different plasma-confinement and heating approaches: the tokamak, the theta-pinch, the magnetic-mirror, and the laser-pellet system. We begin with a discussion of the dynamics of reacting plasmas and basic considerations of reactor power balance. The essential plasma physical aspects of each system are summarized, and the main characteristics of the corresponding conceptual power plants are described. In tokamak reactors the plasma densities are about 1020 m-3, and the ? values (ratio of plasma pressure to confining magnetic pressure) are approximately 5%. Plasma burning times are of the order of 100-1000 sec. Large superconducting dc magnets furnish the toroidal magnetic field, and 2-m thick blankets and shields prevent heat deposition in the superconductor. Radially diffusing plasma is diverted away from the first wall by means of null singularities in the poloidal (or transverse) component of the confining magnetic field. The toroidal theta-pinch reactor has a much smaller minor diameter and a much larger major diameter, and operates on a 10-sec cycle with 0.1-sec burning pulses. It utilizes shock heating from high-voltage sources and adabatic-compression heating powered by low-voltage, pulsed cryogenic magnetic or inertial energy stores, outside the reactor core. The plasma has a density of about 1022 m-3 and ? values of nearly unity. In the power balance of the reactor, direct-conversion energy obtained by expansion of the burning high-? plasma against the containing magnetic field is an important factor. No divertor is necessary since neutral-gas flow cools and replaces the "spent" plasma between pulses. The open-ended mirror reactor uses both thermal conversion of neutron energy and direct conversion of end-loss plasma energy to dc electrical power. A fraction of this direct-convertor power is then fed back to the ioninjection system to sustain the reaction and maintain the plasma. The average ion energy is 600 keV, plasma diameter 6 m, and the plasma beta 85%. The power levels of the three magnetic-confinement devices are in the 500-2000 MWe range, with the exception of the mirror reactor, for which the output is approximately 200 MWe. In Laser-Pellet reactors, frozen D-T pellets are ignited in a cavity which absorbs the electromagnetic, charged particle, and neutron energy from the fusion reaction. The confinement is "inertial," since the fusion reaction occurs during the disassembly of the heated pellet. A pellet-cavity unit would produce about 200 MWt in pulses with a repetition rate of the order of 10 sec-1. Such units could be clustered to give power plants with outputs in the range of 1000 MWe.

F. L. Ribe

1975-01-01T23:59:59.000Z

192

Hybrid adsorptive membrane reactor  

DOE Patents (OSTI)

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

Tsotsis, Theodore T. (Huntington Beach, CA); Sahimi, Muhammad (Altadena, CA); Fayyaz-Najafi, Babak (Richmond, CA); Harale, Aadesh (Los Angeles, CA); Park, Byoung-Gi (Yeosu, KR); Liu, Paul K. T. (Lafayette Hill, PA)

2011-03-01T23:59:59.000Z

193

GEN-IV Reactors  

Science Journals Connector (OSTI)

Generation-IV reactors are a set of nuclear reactors currently being developed under international collaborations targeting ... economics, proliferation resistance, and physical protection of nuclear energy. Nuclear

Taek K. Kim

2013-01-01T23:59:59.000Z

194

The Netherlands Reactor Centre  

Science Journals Connector (OSTI)

... Two illustrated brochures in English have recently J. been issued by the Netherlands Reactor Centre ( ... Centre (Reactor Centrum Nederland). The first* gives a general survey of the ...

S. WEINTROUB

1964-04-04T23:59:59.000Z

195

Routine and post-accident sampling in nuclear reactors  

SciTech Connect

Review of the Three Mile Island accident by NRC has resulted in new post-accident-sampling-capability requirements for utilities that operate pressurized water reactors and/or boiling water reactors. Several vendors are offering equipment that they hope will suffice to met both the new NRC regulations and an operational deadline of January 1, 1981. The advantages and disadvantages of these systems and projected future-new-system needs for TVA reactors are being evaluated in light of TMI experience.

Armento, W.J.; Kitts, F.G.; German, G.E.

1980-01-01T23:59:59.000Z

196

SRS Seals Access to P and R Reactors, Marking End to Nearly 60-Years of  

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

Seals Access to P and R Reactors, Marking End to Nearly Seals Access to P and R Reactors, Marking End to Nearly 60-Years of History: American Recovery and Reinvestment Act Project Progress SRS Seals Access to P and R Reactors, Marking End to Nearly 60-Years of History: American Recovery and Reinvestment Act Project Progress June 28, 2011 - 12:00pm Addthis Marc Sharpe, who was a senior reactor operator at P Reactor in the mid-1980s, (left) and Dr. David Moody, U.S. Department of Energy-Savannah River Operation Office Manager, were the last people to leave P Reactor, just before the opening is welded shut. Marc Sharpe, who was a senior reactor operator at P Reactor in the mid-1980s, (left) and Dr. David Moody, U.S. Department of Energy-Savannah River Operation Office Manager, were the last people to leave P Reactor,

197

Multiobjective Dynamic Optimization of an Industrial Nylon 6 Semibatch Reactor Using Genetic Algorithm  

E-Print Network (OSTI)

Multiobjective Dynamic Optimization of an Industrial Nylon 6 Semibatch Reactor Using Genetic optimization, nylon 6, genetic algorithm, polymer reactor, polymerization, Pareto sets trial nylon 6 reactor for use in the optimal design and operation of these reactors. Some stud- ies4­9 have already been

Coello, Carlos A. Coello

198

High-pressure reaction and emissions characteristics of catalytic reactors for gas turbine combustors  

Science Journals Connector (OSTI)

The reaction and emissions characteristics of catalytic reactors comprising noble metal catalysts were investigated using homogeneous mixtures of natural gas and vitiated air at pressures up to 2.9 MPa. The mixture temperatures at inlet ranged from 500 to 700°C and the fuel-air ratio was increased till the exit gas temperature reached about 1200°C. Values of combustion efficiency greater than 99.5% and nitrogen oxides emissions for all catalytic reactors tested were less than 0.2 g NO2/kg fuel (2 ppm (15% 02) ) for all reactors at reactor exit gas temperatures higher than about 1100°C. Combustion efficiency decreased with increasing pressure in the heterogeneous-reaction controlled region, though a pressure increase favored homogeneous, gas phase reactions. Appreciable reactivity deterioration by aging for 1000 h at 1000°C was observed at lower mixture temperatures. A two-stage combustor comprising a conventional flame combustion stage and a catalytic stage was fabricated and its NO,x emissions and performance were evaluated at conditions typical of stationary gas turbine combustor operations. About 80% reduction in NO,x emissions levels compared with flame combustion was attained at 1 \\{MPa\\} pressure and 1180°C exit gas temperature, together with complete hydrocarbon combustion.

S. Hayashi; H. Yamada; K. Shimodaira

1995-01-01T23:59:59.000Z

199

Reactor noise analysis applications in NPP I and C systems  

SciTech Connect

Reactor noise analysis techniques are used in many NPPs on a routine basis as 'inspection tools' to get information on the dynamics of reactor processes and their instrumentation in a passive, non-intrusive way. The paper discusses some of the tasks and requirements an NPP has to take to implement and to use the full advantages of reactor noise analysis techniques. Typical signal noise analysis applications developed for the monitoring of the reactor shutdown system and control system instrumentation of the Candu units of Ontario Power Generation and Bruce Power are also presented. (authors)

Gloeckler, O. [International Atomic Energy Agency, Wagramer Strosse 5, A-1400 Vienna, Austria Ontario Power Generation, 230 Westney Road South, Ajax, Ont. L1S 7R3 (Canada)

2006-07-01T23:59:59.000Z

200

Immobilization of Fast Reactor First Cycle Raffinate  

SciTech Connect

This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cycle raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method.

Langley, K. F.; Partridge, B. A.; Wise, M.

2003-02-26T23:59:59.000Z

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


201

On the relation between rough set reducts and typical testors  

Science Journals Connector (OSTI)

Abstract This paper studies the relations between rough set reducts and typical testors from the so-called logical combinatorial approach to pattern recognition. Definitions, comments and observations are formally introduced and supported by illustrative examples. Furthermore, some theorems expressing theoretical relations between reducts and typical testors are enunciated and proved. We also discuss several practical applications of these relations that can mutually enrich the development of research and applications in both areas. Although we focus on the relation between the classical concepts of testor and reduct, our study can be expanded to include other types of testors and reducts.

Manuel S. Lazo-Cortés; José Fco. Martínez-Trinidad; Jesús A. Carrasco-Ochoa; Guillermo Sanchez-Diaz

2015-01-01T23:59:59.000Z

202

Finding typical high redshift galaxies with the NOT  

E-Print Network (OSTI)

We present results from an ongoing search for galaxy counterparts of a subgroup of Quasar Absorption Line Systems called Damped Ly-alpha Absorbers (DLAs). DLAs have several characteristics that make them prime candidates for being the progenitors of typical present day galaxies.

J. U. Fynbo; P. Moller; B. Thomsen

1999-12-14T23:59:59.000Z

203

Security Implications of Typical Grid Computing Usage Scenarios Marty Humphrey  

E-Print Network (OSTI)

Security Implications of Typical Grid Computing Usage Scenarios Marty Humphrey Computer Science. A broader goal of these scenarios are to increase the awareness of security issues in Grid Computing. 1 easy and secure ac- cess to the Grid's diverse resources. Infrastructure software such as Legion [6

Thompson, Mary R.

204

Robust Neuroimaging-Based Classification Techniques of Autistic vs. Typically  

E-Print Network (OSTI)

abnormalities in several brain regions. Increased head size was the first observed characteristic in children1 Robust Neuroimaging-Based Classification Techniques of Autistic vs. Typically Developing Brain with autism. According to the published studies, different anatomical structures of the brain have been

Farag, Aly A.

205

SRS Small Modular Reactors  

SciTech Connect

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

None

2012-04-27T23:59:59.000Z

206

SRS Small Modular Reactors  

ScienceCinema (OSTI)

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

None

2014-05-21T23:59:59.000Z

207

University Research Reactor Task Force to the Nuclear Energy Research  

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

University Research Reactor Task Force to the Nuclear Energy University Research Reactor Task Force to the Nuclear Energy Research Advisory Committee University Research Reactor Task Force to the Nuclear Energy Research Advisory Committee In mid-February, 2001 The University Research Reactor (URR) Task Force (TF), a sub-group of the Department of Energy (DOE) Nuclear Energy Research Advisory Committee (NERAC), was asked to: * Analyze information collected by DOE, the NERAC "Blue Ribbon Panel," universities, and other sources pertaining to university reactors including their research and training capabilities, costs to operate, and operating data, and * Provide DOE with clear, near-term recommendations as to actions that should be taken by the Federal Government and a long-term strategy to assure the continued operation of vital university reactor facilities in

208

Cooling system for a nuclear reactor  

DOE Patents (OSTI)

A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

Amtmann, Hans H. (Rancho Santa Fe, CA)

1982-01-01T23:59:59.000Z

209

Nuclear reactor  

DOE Patents (OSTI)

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

Thomson, Wallace B. (Severna Park, MD)

2004-03-16T23:59:59.000Z

210

Corrosion Minimization for Research Reactor Fuel  

SciTech Connect

Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

Eric Shaber; Gerard Hofman

2005-06-01T23:59:59.000Z

211

Light Water Reactor Sustainability (LWRS) Program | Department of Energy  

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

Light Water Reactor Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program The Light Water Reactor Sustainability (LWRS) Program is developing the scientific basis to extend existing nuclear power plant operating life beyond the current 60-year licensing period and ensure long-term reliability, productivity, safety, and security. The program is conducted in collaboration with national laboratories, universities, industry, and international partners. Idaho National Laboratory serves as the Technical Integration Office and coordinates the research and development (R&D) projects in the following pathways: Materials Aging and Degradation Assessment, Advanced Instrumentation, Information, and Control Systems

212

A Study of Fast Reactor Fuel Transmutation in a Candidate Dispersion Fuel Design  

SciTech Connect

Dispersion fuels represent a significant departure from typical ceramic fuels to address swelling and radiation damage in high burnup fuel. Such fuels use a manufacturing process in which fuel particles are encapsulated within a non-fuel matrix. Dispersion fuels have been studied since 1997 as part of an international effort to develop and test very high density fuel types for the Reduced Enrichment for Research and Test Reactors (RERTR) program.[1] The Idaho National Laboratory is performing research in the development of an innovative dispersion fuel concept that will meet the challenges of transuranic (TRU) transmutation by providing an integral fission gas plenum within the fuel itself, to eliminate the swelling that accompanies the irradiation of TRU. In this process, a metal TRU vector produced in a separations process is atomized into solid microspheres. The dispersion fuel process overcoats the microspheres with a mixture of resin and hollow carbon microspheres to create a TRUC. The foam may then be heated and mixed with a metal power (e.g., Zr, Ti, or Si) and resin to form a matrix metal carbide, that may be compacted and extruded into fuel elements. In this paper, we perform reactor physics calculations for a core loaded with the conceptual fuel design. We will assume a “typical” TRU vector and a reference matrix density. We will employ a fuel and core design based on the Advanced Burner Test Reactor (ABTR) design.[2] Using the CSAS6 and TRITON modules of the SCALE system [3] for preliminary scoping studies, we will demonstrate the feasibility of reactor operations. This paper will describe the results of these analyses.

Mark DeHart; Hongbin Zhang; Eric Shaber; Matthew Jesse

2010-11-01T23:59:59.000Z

213

Maximum Photovoltaic Penetration Levels on Typical Distribution Feeders: Preprint  

SciTech Connect

This paper presents simulation results for a taxonomy of typical distribution feeders with various levels of photovoltaic (PV) penetration. For each of the 16 feeders simulated, the maximum PV penetration that did not result in steady-state voltage or current violation is presented for several PV location scenarios: clustered near the feeder source, clustered near the midpoint of the feeder, clustered near the end of the feeder, randomly located, and evenly distributed. In addition, the maximum level of PV is presented for single, large PV systems at each location. Maximum PV penetration was determined by requiring that feeder voltages stay within ANSI Range A and that feeder currents stay within the ranges determined by overcurrent protection devices. Simulations were run in GridLAB-D using hourly time steps over a year with randomized load profiles based on utility data and typical meteorological year weather data. For 86% of the cases simulated, maximum PV penetration was at least 30% of peak load.

Hoke, A.; Butler, R.; Hambrick, J.; Kroposki, B.

2012-07-01T23:59:59.000Z

214

Microsoft Word - illinois_reactors_taiwo.doc  

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

Fission Process and Control Fission Process and Control In nuclear power reactors, energy is produced by the nuclear fission process in which uranium atoms are split into two major atoms, called fission products, with significant heat generation. A nuclear reactor system is controlled to ensure that the fission process is a sustained nuclear chain reaction (see Fig. 1) that neither declines nor increases with operation time, i.e., it is at

215

Tokamak fusion reactors with less than full tritium breeding  

SciTech Connect

A study of commercial, tokamak fusion reactors with tritium concentrations and tritium breeding ratios ranging from full deuterium-tritium operation to operation with no tritium breeding is presented. The design basis for these reactors is similar to those of STARFIRE and WILDCAT. Optimum operating temperatures, sizes, toroidal field strengths, and blanket/shield configurations are determined for a sequence of reactor designs spanning the range of tritium breeding, each having the same values of beta, thermal power, and first-wall heat load. Additional reactor parameters, tritium inventories and throughputs, and detailed costs are calculated for each reactor design. The disadvantages, advantages, implications, and ramifications of tritium-depleted operation are presented and discussed.

Evans, K. Jr.; Gilligan, J.G.; Jung, J.

1983-05-01T23:59:59.000Z

216

Assessing low power reactivity levels in subcritical CANDU reactors  

SciTech Connect

A new technique has been developed for monitoring slow reactivity changes in Canadian deuterium uranium (CANDU) nuclear reactors during low power operation, following a sustained period of high power operation. The power doubling halving worth (PDHW) test tracks slow reactivity changes by evaluating the reactivity perturbation required by the power regulating system to halve and double reactor power over time. During low power operation of a CANDU reactor, the PDHW test is used to monitor the decay of the photoneutron precursors, so that the reactor power can be lowered using a preset amount of reactivity. The PDHW test is described in this paper and is validated by using computer simulations and operating data from a CANDU reactor.

Teare, S.W. [Corporation of the City of Mississauga, Ontario (Canada)] [Corporation of the City of Mississauga, Ontario (Canada); Shanes, F.C. [Ontario Hydro, Toronto, Ontario (Canada)] [Ontario Hydro, Toronto, Ontario (Canada); Hersey, M.W. [Pickering Nuclear Generating Station, Ontario (Canada)] [Pickering Nuclear Generating Station, Ontario (Canada)

1996-08-01T23:59:59.000Z

217

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

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

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

218

Dual bed reactor for the study of catalytic biomass tars conversion  

SciTech Connect

A dual fixed bed laboratory scale set up has been used to compare the activity of a novel Rh/LaCoO{sub 3}/Al{sub 2}O{sub 3} catalyst to that of dolomite, olivine and Ni/Al{sub 2}O{sub 3}, typical catalysts used in fluidized bed biomass gasification, to convert tars produced during biomass devolatilization stage. The experimental apparatus allows the catalyst to be operated under controlled conditions of temperature and with a real gas mixture obtained by the pyrolysis of the biomass carried out in a separate fixed bed reactor operated under a selected and controlled heating up rate. The proposed catalyst exhibits much better performances than conventional catalysts tested. It is able to completely convert tars and also to strongly decrease coke formation due to its good redox properties. (author)

Ammendola, P.; Piriou, B.; Lisi, L.; Ruoppolo, G.; Chirone, R.; Russo, G. [Istituto di Ricerche sulla Combustione - CNR, P.le V. Tecchio 80, 80125 Napoli (Italy)

2010-04-15T23:59:59.000Z

219

Metal fires and their implications for advanced reactors.  

SciTech Connect

This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in these areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety analysis capabilities of the advanced-reactor community for directly relevant scenarios. Beyond the focus on the thermally-interacting and smoldering sodium pool fires, experimental and analysis capabilities for sodium spray fires have also been developed in this project.

Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean; Hewson, John C.; Blanchat, Thomas K.

2010-10-01T23:59:59.000Z

220

A Computuerized Operator Support System Prototype  

SciTech Connect

A report was published by the Idaho National Laboratory in September of 2012, entitled Design to Achieve Fault Tolerance and Resilience, which described the benefits of automating operator actions for transients. The report identified situations in which providing additional automation in lieu of operator actions would be advantageous. It recognized that managing certain plant upsets is sometimes limited by the operator’s ability to quickly diagnose the fault and to take the needed actions in the time available. Undoubtedly, technology is underutilized in the nuclear power industry for operator assistance during plant faults and operating transients. In contrast, other industry sectors have amply demonstrated that various forms of operator advisory systems can enhance operator performance while maintaining the role and responsibility of the operator as the independent and ultimate decision-maker. A computerized operator support system (COSS) is proposed for use in nuclear power plants to assist control room operators in addressing time-critical plant upsets. A COSS is a collection of technologies to assist operators in monitoring overall plant performance and making timely, informed decisions on appropriate control actions for the projected plant condition. The COSS does not supplant the role of the operator, but rather provides rapid assessments, computations, and recommendations to reduce workload and augment operator judgment and decision-making during fast-moving, complex events. This project proposes a general model for a control room COSS that addresses a sequence of general tasks required to manage any plant upset: detection, validation, diagnosis, recommendation, monitoring, and recovery. The model serves as a framework for assembling a set of technologies that can be interrelated to assist with each of these tasks. A prototype COSS has been developed in order to demonstrate the concept and provide a test bed for further research. The prototype is based on four underlying elements consisting of a digital alarm system, computer-based procedures, PI&D system representations, and a recommender module for mitigation actions. At this point, the prototype simulates an interface to a sensor validation module and a fault diagnosis module. These two modules will be fully integrated in the next version of the prototype. The initial version of the prototype is now operational at the Idaho National Laboratory using the U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) Human Systems Simulation Laboratory (HSSL). The HSSL is a full-scope, full-scale glass top simulator capable of simulating existing and future nuclear power plant main control rooms. The COSS is interfaced to the Generic Pressurized Water Reactor (gPWR) simulator with industry-typical control board layouts. The glass top panels display realistic images of the control boards that can be operated by touch gestures. A section of the simulated control board was dedicated to the COSS human-system interface (HSI), which resulted in a seamless integration of the COSS into the normal control room environment.

Ken Thomas; Ronald Boring; Roger Lew; Tom Ulrich; Richard Villim

2013-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor operators typically" 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

A Computuerized Operator Support System Prototype  

SciTech Connect

A report was published by the Idaho National Laboratory in September of 2012, entitled Design to Achieve Fault Tolerance and Resilience, which described the benefits of automating operator actions for transients. The report identified situations in which providing additional automation in lieu of operator actions would be advantageous. It recognized that managing certain plant upsets is sometimes limited by the operator’s ability to quickly diagnose the fault and to take the needed actions in the time available. Undoubtedly, technology is underutilized in the nuclear power industry for operator assistance during plant faults and operating transients. In contrast, other industry sectors have amply demonstrated that various forms of operator advisory systems can enhance operator performance while maintaining the role and responsibility of the operator as the independent and ultimate decision-maker. A computerized operator support system (COSS) is proposed for use in nuclear power plants to assist control room operators in addressing time-critical plant upsets. A COSS is a collection of technologies to assist operators in monitoring overall plant performance and making timely, informed decisions on appropriate control actions for the projected plant condition. The COSS does not supplant the role of the operator, but rather provides rapid assessments, computations, and recommendations to reduce workload and augment operator judgment and decision-making during fast-moving, complex events. This project proposes a general model for a control room COSS that addresses a sequence of general tasks required to manage any plant upset: detection, validation, diagnosis, recommendation, monitoring, and recovery. The model serves as a framework for assembling a set of technologies that can be interrelated to assist with each of these tasks. A prototype COSS has been developed in order to demonstrate the concept and provide a test bed for further research. The prototype is based on four underlying elements consisting of a digital alarm system, computer-based procedures, PI&D system representations, and a recommender module for mitigation actions. At this point, the prototype simulates an interface to a sensor validation module and a fault diagnosis module. These two modules will be fully integrated in the next version of the prototype. The initial version of the prototype is now operational at the Idaho National Laboratory using the U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) Human Systems Simulation Laboratory (HSSL). The HSSL is a full-scope, full-scale glass top simulator capable of simulating existing and future nuclear power plant main control rooms. The COSS is interfaced to the Generic Pressurized Water Reactor (gPWR) simulator with industry-typical control board layouts. The glass top panels display realistic images of the control boards that can be operated by touch gestures. A section of the simulated control board was dedicated to the COSS human-system interface (HSI), which resulted in a seamless integration of the COSS into the normal control room environment.

Ken Thomas; Ronald Boring; Roger Lew; Tom Ulrich; Richard Villim

2013-11-01T23:59:59.000Z

222

Is the Sun Embedded in a Typical Interstellar Cloud?  

E-Print Network (OSTI)

The physical properties and kinematics of the partially ionized interstellar material near the Sun are typical of warm diffuse clouds in the solar vicinity. The interstellar magnetic field at the heliosphere and the kinematics of nearby clouds are naturally explained in terms of the S1 superbubble shell. The interstellar radiation field at the Sun appears to be harder than the field ionizing ambient diffuse gas, which may be a consequence of the low opacity of the tiny cloud surrounding the heliosphere. The spatial context of the Local Bubble is consistent with our location in the Orion spur.

P. C. Frisch

2008-04-23T23:59:59.000Z

223

High Flux Beam Reactor | Environmental Restoration Projects | BNL  

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

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

224

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

225

Elementary Reactor Physics  

Science Journals Connector (OSTI)

... THERE are few subjects which have developed at the rate at which reactor physics and ... physics and reactor theory have done. This, of course, is largely due to the circumstances in ...

J. F. HILL

1962-02-10T23:59:59.000Z

226

Colliding Beam Fusion Reactors  

Science Journals Connector (OSTI)

The recirculating power for virtually all types of fusion reactors has previously been calculated [1] with the Fokker–Planck equation. The reactors involve non-Maxwellian plasmas. The calculations are ... the rec...

Norman Rostoker; Artan Qerushi; Michl Binderbauer

2003-06-01T23:59:59.000Z

227

Development of the Mathematics of Learning Curve Models for Evaluating Small Modular Reactor Economics  

SciTech Connect

The cost of nuclear power is a straightforward yet complicated topic. It is straightforward in that the cost of nuclear power is a function of the cost to build the nuclear power plant, the cost to operate and maintain it, and the cost to provide fuel for it. It is complicated in that some of those costs are not necessarily known, introducing uncertainty into the analysis. For large light water reactor (LWR)-based nuclear power plants, the uncertainty is mainly contained within the cost of construction. The typical costs of operations and maintenance (O&M), as well as fuel, are well known based on the current fleet of LWRs. However, the last currently operating reactor to come online was Watts Bar 1 in May 1996; thus, the expected construction costs for gigawatt (GW)-class reactors in the United States are based on information nearly two decades old. Extrapolating construction, O&M, and fuel costs from GW-class LWRs to LWR-based small modular reactors (SMRs) introduces even more complication. The per-installed-kilowatt construction costs for SMRs are likely to be higher than those for the GW-class reactors based on the property of the economy of scale. Generally speaking, the economy of scale is the tendency for overall costs to increase slower than the overall production capacity. For power plants, this means that doubling the power production capacity would be expected to cost less than twice as much. Applying this property in the opposite direction, halving the power production capacity would be expected to cost more than half as much. This can potentially make the SMRs less competitive in the electricity market against the GW-class reactors, as well as against other power sources such as natural gas and subsidized renewables. One factor that can potentially aid the SMRs in achieving economic competitiveness is an economy of numbers, as opposed to the economy of scale, associated with learning curves. The basic concept of the learning curve is that the more a new process is repeated, the more efficient the process can be made. Assuming that efficiency directly relates to cost means that the more a new process is repeated successfully and efficiently, the less costly the process can be made. This factor ties directly into the factory fabrication and modularization aspect of the SMR paradigm—manufacturing serial, standardized, identical components for use in nuclear power plants can allow the SMR industry to use the learning curves to predict and optimize deployment costs.

Harrison, T. J. [ORNL

2014-02-01T23:59:59.000Z

228

Meteorology: typical meteorological year data for selected stations in  

Open Energy Info (EERE)

Ethiopia from NREL Ethiopia from NREL Dataset Summary Description (Abstract): Each TMY is a data set of hourly values of solar radiation and meteorological elements for a 1-year period. Solar radiation is modeled using the NREL METSTAT model, with surface observed cloud cover being the principal model input. The container file contains one TMY file for each selected station in the region, plus documentation files and a TMY data reader file for use with Microsoft Excel. (Purpose): Simulations (Supplemental Information): A TMY consists of months selected from individual years and concatenated to form a complete year. The intended use is for computer simulations of solar energy conversion systems and building systems. Because of the selection criteria, these TMYs are not appropriate for simulations of wind energy conversion systems. A TMY provides a standard for hourly data for solar radiation and other meteorological elements that permit performance comparisons of system types and configurations for one or more locations. A TMY is not necessarily a good indicator of conditions over the next year, or even the next 5 years. Rather, it represents conditions judged to be typical over a long period of time, such as 30 years. Because they represent typical rather than extreme conditions, they are not suited for designing systems and their components to meet the worst-case conditions

229

Meteorology: typical meteorological year data for selected stations in  

Open Energy Info (EERE)

Brazil from NREL Brazil from NREL Dataset Summary Description (Abstract): Each TMY is a data set of hourly values of solar radiation and meteorological elements for a 1-year period. Solar radiation is modeled using the NREL METSTAT model, with surface observed cloud cover being the principal model input. The container file contains one TMY file for each selected station in the region, plus documentation files and a TMY data reader file for use with Microsoft Excel. (Purpose): Simulations (Supplemental Information): A TMY consists of months selected from individual years and concatenated to form a complete year. The intended use is for computer simulations of solar energy conversion systems and building systems. Because of the selection criteria, these TMYs are not appropriate for simulations of wind energy conversion systems. A TMY provides a standard for hourly data for solar radiation and other meteorological elements that permit performance comparisons of system types and configurations for one or more locations. A TMY is not necessarily a good indicator of conditions over the next year, or even the next 5 years. Rather, it represents conditions judged to be typical over a long period of time, such as 30 years. Because they represent typical rather than extreme conditions, they are not suited for designing systems and their components to meet the worst-case conditions occurring at a location.

230

Meteorology: typical meteorological year data for selected stations in  

Open Energy Info (EERE)

Nepal from NREL Nepal from NREL Dataset Summary Description (Abstract): Each TMY is a data set of hourly values of solar radiation and meteorological elements for a 1-year period. Solar radiation is modeled using the NREL METSTAT model, with surface observed cloud cover being the principal model input. The container file contains one TMY file for each selected station in the region, plus documentation files and a TMY data reader file for use with Microsoft Excel. (Purpose): Simulations (Supplemental Information): A TMY consists of months selected from individual years and concatenated to form a complete year. The intended use is for computer simulations of solar energy conversion systems and building systems. Because of the selection criteria, these TMYs are not appropriate for simulations of wind energy conversion systems. A TMY provides a standard for hourly data for solar radiation and other meteorological elements that permit performance comparisons of system types and configurations for one or more locations. A TMY is not necessarily a good indicator of conditions over the next year, or even the next 5 years. Rather, it represents conditions judged to be typical over a long period of time, such as 30 years. Because they represent typical rather than extreme conditions, they are not suited for designing systems and their components to meet the worst-case conditions occurring at a location.

231

Meteorology: typical meteorological data for selected stations in Ghana  

Open Energy Info (EERE)

data for selected stations in Ghana data for selected stations in Ghana from NREL Dataset Summary Description (Abstract): Each TMY is a data set of hourly values of solar radiation and meteorological elements for a 1-year period. Solar radiation is modeled using the NREL METSTAT model, with surface observed cloud cover being the principal model input. The container file contains one TMY file for each selected station in the region, plus documentation files and a TMY data reader file for use with Microsoft Excel. (Purpose): Simulations> (Supplemental Information): A TMY consists of months selected from individual years and concatenated to form a complete year. The intended use is for computer simulations of solar energy conversion systems and building systems. Because of the selection criteria, these TMYs are not appropriate for simulations of wind energy conversion systems. A TMY provides a standard for hourly data for solar radiation and other meteorological elements that permit performance comparisons of system types and configurations for one or more locations. A TMY is not necessarily a good indicator of conditions over the next year, or even the next 5 years. Rather, it represents conditions judged to be typical over a long period of time, such as 30 years. Because they represent typical rather than extreme conditions, they are not suited for designing systems and their components to meet the worst-case conditions occurring at a location.

232

Meteorology: typical meteorological year data for selected stations in Sri  

Open Energy Info (EERE)

Sri Sri Lanka from NREL Dataset Summary Description (Abstract): A data set of hourly values of solar radiation and meteorological elements for a 1-year period. (Purpose): Simulations (Supplemental Information): A TMY consists of months selected from individual years and concatenated to form a complete year. The intended use is for computer simulations of solar energy conversion systems and building systems. Because of the selection criteria, these TMYs are not appropriate for simulations of wind energy conversion systems. A TMY provides a standard for hourly data for solar radiation and other meteorological elements that permit performance comparisons of system types and configurations for one or more locations. A TMY is not necessarily a good indicator of conditions over the next year, or even the next 5 years. Rather, it represents conditions judged to be typical over a long period of time, such as 30 years. Because they represent typical rather than extreme conditions, they are not suited for designing systems and their components to meet the worst-case conditions occurring at a location.

233

Meteorology: typical meteorological year data for selected stations in  

Open Energy Info (EERE)

Kenya from NREL Kenya from NREL Dataset Summary Description (Abstract): Each TMY is a data set of hourly values of solar radiation and meteorological elements for a 1-year period. Solar radiation is modeled using the NREL METSTAT model, with surface observed cloud cover being the principal model input. The container file contains one TMY file for each selected station in the region, plus documentation files and a TMY data reader file for use with Microsoft Excel. (Purpose): Simulations (Supplemental Information): A TMY consists of months selected from individual years and concatenated to form a complete year. The intended use is for computer simulations of solar energy conversion systems and building systems. Because of the selection criteria, these TMYs are not appropriate for simulations of wind energy conversion systems. A TMY provides a standard for hourly data for solar radiation and other meteorological elements that permit performance comparisons of system types and configurations for one or more locations. A TMY is not necessarily a good indicator of conditions over the next year, or even the next 5 years. Rather, it represents conditions judged to be typical over a long period of time, such as 30 years. Because they represent typical rather than extreme conditions, they are not suited for designing systems and their components to meet the worst-case conditions occurring at a location.

234

Materials Degradation in Light Water Reactors: Life After 60 | Department  

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

Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the

235

Brookhaven Graphite Research Reactor Workshop | Department of Energy  

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

Services » Site & Facility Restoration » Deactivation & Services » Site & Facility Restoration » Deactivation & Decommissioning (D&D) » D&D Workshops » Brookhaven Graphite Research Reactor Workshop Brookhaven Graphite Research Reactor Workshop The Brookhaven Graphite Research Reactor (BGRR) was the first reactor built in the U.S. for peacetime atomic research following World War II. Construction began in 1947 and the reactor started operating in August 1950. In the next 18 years, an estimated 25,000 scientific experiments were carried out at the BGRR using neutrons produced in the facility's 700-ton graphite core, made up of more than 60,000 individual graphite blocks. The BGRR was placed on standby in 1968 and then permanently shut down as the next-generation reactor, the High Flux Beam Reactor (HFBR), was

236

Manhattan Project: Production Reactor (Pile) Design, Met Lab, 1942  

Office of Scientific and Technical Information (OSTI)

Schematic of the X-10 Graphite Reactor, Oak Ridge PRODUCTION REACTOR (PILE) DESIGN Schematic of the X-10 Graphite Reactor, Oak Ridge PRODUCTION REACTOR (PILE) DESIGN (Met Lab, 1942) 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 By 1942, scientists had established that some of the uranium exposed to radioactivity in a reactor (pile) would eventually decay into plutonium, which could then be separated by chemical means from the uranium. Important theoretical research on this was ongoing, but the work was scattered at various universities from coast to coast. In early 1942, Arthur Compton arranged for all pile research to be moved to the Met Lab at the University of Chicago.

237

Materials Degradation in Light Water Reactors: Life After 60 | Department  

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

Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the

238

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research  

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

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program The Department of Energy's (DOE's) Light Water Reactor Sustainability (LWRS) Program is a five year effort that works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operation of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging

239

A Methodology for the Neutronics Design of Space Nuclear Reactors  

SciTech Connect

A methodology for the neutronics design of space power reactors is presented. This methodology involves balancing the competing requirements of having sufficient excess reactivity for the desired lifetime, keeping the reactor subcritical at launch and during submersion accidents, and providing sufficient control over the lifetime of the reactor. These requirements are addressed by three reactivity values for a given reactor design: the excess reactivity at beginning of mission, the negative reactivity at shutdown, and the negative reactivity margin in submersion accidents. These reactivity values define the control worth and the safety worth in submersion accidents, used for evaluating the merit of a proposed reactor type and design. The Heat Pipe-Segmented Thermoelectric Module Converters space reactor core design is evaluated and modified based on the proposed methodology. The final reactor core design has sufficient excess reactivity for 10 years of nominal operation at 1.82 MW of fission power and is subcritical at launch and in all water submersion accidents.

King, Jeffrey C.; El-Genk, Mohamed S. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM 87131 (United States); Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM 87131 (United States)

2004-02-04T23:59:59.000Z

240

Prospects for spheromak fusion reactors  

Science Journals Connector (OSTI)

The reactor study of Hagenson and Krakowski demonstrated the attractiveness of the spheromak as a compact fusion reactor, based on...

T. K. Fowler; D. D. Hua

1995-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor operators typically" 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

Advanced Test Reactor Tour  

SciTech Connect

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

Miley, Don

2011-01-01T23:59:59.000Z

242

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

243

Advanced Test Reactor Tour  

ScienceCinema (OSTI)

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

Miley, Don

2013-05-28T23:59:59.000Z

244

Conduct of Operations Criteria, Review, & Approach Documents | Department  

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

Conduct of Operations Criteria, Review, & Approach Documents Conduct of Operations Criteria, Review, & Approach Documents Conduct of Operations Criteria, Review, & Approach Documents Documents Available for Download CRAD, Conduct of Operations - Idaho MF-628 Drum Treatment Facility CRAD, Conduct of Operations - Idaho Accelerated Retrieval Project Phase II CRAD, Conduct of Operations - Los Alamos National Laboratory TA 55 SST Facility CRAD, Conduct of Operations - Los Alamos National Laboratory Waste Characterization, Reduction, and Repackaging Facility CRAD, Conduct of Operations - Oak Ridge National Laboratory High Flux Isotope Reactor CRAD, Conduct of Operations - Oak Ridge National Laboratory High Flux Isotope Reactor Contractor ORR CRAD, Conduct of Operations - Oak Ridge National Laboratory TRU ALPHA LLWT Project

245

Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies  

SciTech Connect

A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

Dixon, David D. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Hiatt, Matthew T. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering, Texas A and M University, College Station, TX 77843 (United States); Poston, David I.; Kapernick, Richard J. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

2006-01-20T23:59:59.000Z

246

A comparison of factors impacting on radiation buildup at the Vermont Yankee and Monticello BWRs (boiling-water reactors): Interim report  

SciTech Connect

Design and operating features of the Monticello and Vermont Yankee BWRs were compared in an attempt to explain why shutdown radiation levels at Vermont Yankee were significantly higher than at Monticello. The plants were shown to be similar in many respects, for example, condenser and feedwater system design and materials, condensate treatment system design, feedwater iron and copper concentrations, reactor water piping materials and fabrication techniques, reactor water cleanup system flowrates and equipment type, fuel cycle lengths, and fuel failure history. Differences were noted in core power density, jet pump design, reactor water conductivity, volume of radwaste recycle, and the amount of Stellite bearing materials in the feedwater system. Corrosion films on reactor system decontamination flanges from the two plants also were very different. At Monticello, the film was typical of that observed at other BWRs. The Vermont Yankee film contained significantly higher levels of zinc, chromium, and cobalt. Since reactor water Co-60 concentrations at Monticello were about twice those at Vermont Yankee, the Vermont Yankee corrosion film must exhibit a greater tendency to incorporate Co-60.

Palino, G.F.; Hobart, R.L.; Sawochka, S.G.

1987-03-01T23:59:59.000Z

247

Design options for a bunsen reactor.  

SciTech Connect

This work is being performed for Matt Channon Consulting as part of the Sandia National Laboratories New Mexico Small Business Assistance Program (NMSBA). Matt Channon Consulting has requested Sandia's assistance in the design of a chemical Bunsen reactor for the reaction of SO2, I2 and H2O to produce H2SO4 and HI with a SO2 feed rate to the reactor of 50 kg/hour. Based on this value, an assumed reactor efficiency of 33%, and kinetic data from the literature, a plug flow reactor approximately 1%E2%80%9D diameter and and 12 inches long would be needed to meet the specification of the project. Because the Bunsen reaction is exothermic, heat in the amount of approximately 128,000 kJ/hr would need to be removed using a cooling jacket placed around the tubular reactor. The available literature information on Bunsen reactor design and operation, certain support equipment needed for process operation and a design that meet the specification of Matt Channon Consulting are presented.

Moore, Robert Charles

2013-10-01T23:59:59.000Z

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

Predicting aerodynamic characteristic of typical wind turbine airfoils using CFD  

SciTech Connect

An investigation was conducted into the capabilities and accuracy of a representative computational fluid dynamics code to predict the flow field and aerodynamic characteristics of typical wind-turbine airfoils. Comparisons of the computed pressure and aerodynamic coefficients were made with wind tunnel data. This work highlights two areas in CFD that require further investigation and development in order to enable accurate numerical simulations of flow about current generation wind-turbine airfoils: transition prediction and turbulence modeling. The results show that the laminar-to turbulent transition point must be modeled correctly to get accurate simulations for attached flow. Calculations also show that the standard turbulence model used in most commercial CFD codes, the k-e model, is not appropriate at angles of attack with flow separation. 14 refs., 28 figs., 4 tabs.

Wolfe, W.P. [Sandia National Labs., Albuquerque, NM (United States); Ochs, S.S. [Iowa State Univ., Ames, IA (United States). Aerospace Engineering Dept.

1997-09-01T23:59:59.000Z

250

The Sun. A typical star in the solar neighborhood?  

E-Print Network (OSTI)

The Sun is used as the fundamental standard in chemical abundance studies, thus it is important to know whether the solar abundance pattern is representative of the solar neighborhood. Albeit at low precision (0.05 - 0.10 dex) the Sun seems to be a typical solar-metallicity disk star, at high precision (0.01 dex) its abundance pattern seems abnormal when compared to solar twins. The Sun shows a deficiency of refractory elements that could be due to the formation of terrestrial planets. The formation of giant planets may also introduce a signature in the chemical composition of stars. We discuss both planet signatures and also the enhancement of neutron-capture elements in the solar twin 18 Sco.

Melendez, Jorge

2013-01-01T23:59:59.000Z

251

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

252

Advanced Test Reactor National Scientific User Facility  

SciTech Connect

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

253

Design, construction and evaluation of a facility for the simulation of fast reactor blankets  

E-Print Network (OSTI)

A facility has been designed and constructed at the MIT Reactor for the experimental investigation of typical LMFBR breeding blankets. A large converter assembly, consisting of a 20-cm-thick layer of graphite followed by ...

Forbes, Ian Alexander

1970-01-01T23:59:59.000Z

254

A regression approach for Zircaloy-2 in-reactor creep constitutive equations  

E-Print Network (OSTI)

Typical data analysis procedures used in developing current phenomenological in-reactor creep equations for Zircaloy cladding materials are examined. It is found that the data normalization assumptions and the curve fitting ...

Liu, Yung-Yuan

255

Photo of the Week: The Sixth Zero Power Reactor | Department of Energy  

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

The Sixth Zero Power Reactor The Sixth Zero Power Reactor Photo of the Week: The Sixth Zero Power Reactor November 13, 2013 - 4:41pm Addthis In 1958, Argonne National Laboratory began the construction of several zero-power reactors (ZPRs), which are nuclear fission reactors that don't actually generate any power. Scientists developed ZPRs to assess the performance of various reactor core configurations before actually building a full nuclear reactor. A series of ZPRs were built leading up to the construction of the Experimental Breeder Reactor-II, a sodium-cooled fast reactor power plant. In this 1970 photo, an Argonne scientist is loading the matrices of the ZPR-VI reactor prior to its first operation using plutonium fuel. | Photo courtesy of the Department of Energy. In 1958, Argonne National Laboratory began the construction of several

256

Spectral Structure of Electron Antineutrinos from Nuclear Reactors  

E-Print Network (OSTI)

Recent measurements of the positron energy spectrum obtained from inverse beta decay interactions of reactor electron antineutrinos show an excess in the 4 to 6 MeV region relative to current predictions. First-principle calculations of fission and beta decay processes within a typical pressurized water reactor core identify prominent fission daughter isotopes as a possible origin for this excess. These calculations also predict percent-level substructure in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of this substructure can constrain nuclear reactor physics. The substructure can be a systematic uncertainty for measurements utilizing the detailed spectral shape.

D. A. Dwyer; T. J. Langford

2014-07-04T23:59:59.000Z

257

NUCLEAR POWER AND RESEARCH REACTORS 1939 1942 1943 1944  

E-Print Network (OSTI)

#12;#12;11 #12;2 NUCLEAR POWER AND RESEARCH REACTORS 1939 1942 1943 1944 Nuclear fission discovered Oak Ridge selected as site for World War II Manhattan Project First sustained and controlled nuclear 430 nuclear power reactors are operating in the world, and 103 nuclear power plants produce 20

Pennycook, Steve

258

Inherently Safe Reactors and a Second Nuclear Era  

Science Journals Connector (OSTI)

...system pumps and the steam generators are also...of active safety systems or reactor operators...must be proven. Tools and proce-dures...must be devised. Steam generators must be...deploy a new reactor system such as PIUS or even...proba-bilistic risk assessment, is perhaps 10 to...

Alvin M. Weinberg; Irving Spiewak

1984-06-29T23:59:59.000Z

259

Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors  

SciTech Connect

Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

2005-10-01T23:59:59.000Z

260

Westinghouse Small Modular Reactor balance of plant and supporting systems design  

SciTech Connect

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

Memmott, M. J.; Stansbury, C.; Taylor, C. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

2012-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor operators typically" 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

Modeling of membrane reactor for steam methane reforming: From granular to structured catalysts  

Science Journals Connector (OSTI)

Different types and operating modes of a tubular membrane reactor for steam methane reforming with a production rate of 0.6...

A. B. Shigarov; V. A. Kirillov

2012-04-01T23:59:59.000Z

262

E-Print Network 3.0 - application research reactor Sample Search...  

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

Nuclear Technologies 28 Research Aptitude Problem 1 Scavenging of aerosol particles by ice crystals Summary: strategies that would be required to operate these reactor systems....

263

EVALUATION OF ACTIVATION PRODUCTS IN REMAINING IN REMAINING K-, L- AND C-REACTOR STRUCTURES  

SciTech Connect

An analytic model and calculational methodology was previously developed for P-reactor and R-reactor to quantify the radioisotopes present in Savannah River Site (SRS) reactor tanks and the surrounding structural materials as a result of neutron activation of the materials during reactor operation. That methodology has been extended to K-reactor, L-reactor, and C-reactor. The analysis was performed to provide a best-estimate source term input to the Performance Assessment for an in-situ disposition strategy by Site Decommissioning and Demolition (SDD). The reactor structure model developed earlier for the P-reactor and R-reactor analyses was also used for the K-reactor and L-reactor. The model was suitably modified to handle the larger Creactor tank and associated structures. For all reactors, the structure model consisted of 3 annular zones, homogenized by the amount of structural materials in the zone, and 5 horizontal layers. The curie content on an individual radioisotope basis and total basis for each of the regions was determined. A summary of these results are provided herein. The efficacy of this methodology to accurately predict the radioisotopic content of the reactor systems in question has been demonstrated and is documented in Reference 1. As noted in that report, results for one reactor facility cannot be directly extrapolated to other SRS reactors.

Vinson, D.; Webb, R.

2010-09-30T23:59:59.000Z

264

Minor Actinide Transmutation Potential of Modified PROMETHEUS Fusion Reactor  

Science Journals Connector (OSTI)

This study presents the investigation of the burning and/or transmutation (B/T) of minor actinides (MAs) in the modified PROMETHEUS-H fusion reactor. The calculations were performed for an operation...2.... In or...

Hüseyin Yap??c??; Gamze Genç; Nesrin Demir; Bilge Çeper

2004-06-01T23:59:59.000Z

265

Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations  

SciTech Connect

Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated cost of decommissioning a PWR is lowest for ENTOMB and highest for SAFSTOR • the estimated cost of decommissioning a BWR is lowest for OECON and highest for SAFSTOR. In all cases, SAFSTOR has the lowest occupational radiation dose and the highest cost.

Wittenbrock, N. G.

1982-01-01T23:59:59.000Z

266

Safety of Department of Energy-Owned Nuclear Reactors  

Directives, Delegations, and Requirements

To establish reactor safety program requirements assure that the safety of each Department of Energy-owned (DOE-owned) reactor is properly analyzed, evaluated, documented, and approved by DOE; and reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate protection for health and safety and will be in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. Cancels Chap. 6 of DOE O 5480.1A. Paragraphs 7b(3), 7e(3) & 8c canceled by DOE O 5480.23 & canceled by DOE N 251.4 of 9-29-95.

1986-09-23T23:59:59.000Z

267

Savannah River Site K-Reactor Probabilistic Safety Assessment  

SciTech Connect

This report gives the results of a Savannah River Site (SRS) K-Reactor Probabilistic Safety Assessment (PSA). Measures of adverse consequences to health and safety resulting from representations of severe accidents in SRS reactors are presented. In addition, the report gives a summary of the methods employed to represent these accidents and to assess the resultant consequences. The report is issued to provide useful information to the U. S. Department of Energy (DOE) on the risk of operation of SRS reactors, for insights into severe accident phenomena that contribute to this risk, and in support of improved bases for other DOE programs in Heavy Water Reactor safety.

Brandyberry, M.D.; Bailey, R.T.; Baker, W.H.; Kearnaghan, D.P.; O`Kula, K.R.; Wittman, R.S.; Woody, N.D. [Westinghouse Savannah River Co., Aiken, SC (United States); Amos, C.N.; Weingardt, J.J. [Science Applications International Corp. (United States)

1992-12-01T23:59:59.000Z

268

Gaseous fission product management for molten salt reactors and vented fuel systems  

SciTech Connect

Fission gas disposal is one of the unresolved difficulties for Molten Salt Reactors (MSRs) and advanced reactors with vented fuel systems. As these systems operate, they produce many radioactive isotopes of xenon and krypton (e.g. {sup 135}Xe t{sub 1/2} = 9.14 hours and {sup 85}Kr t{sub 1/2}= 10.73 years). Removing these gases proves vital to the success of such reactor designs for two reasons. First, the gases act as large neutron sinks which decrease reactivity and must be counterbalanced by increasing fuel loading. Second, for MSRs, inert fission product gases naturally separate quickly from high temperature salts, thus creating high vapor pressure which poses safety concerns. For advanced reactors with solid vented fuel, the gases are allowed to escape into an off-gas system and thus must be managed. Because of time delays in transport of fission product gases in vented fuel systems, some of the shorter-lived radionuclides will decay away thereby reducing the fission gas source term relative to an MSR. To calculate the fission gas source term of a typical molten salt reactor, we modeled a 1000 MWe graphite moderated thorium MSR similar to that detailed in Mathieu et al. [1]. The fuel salt used in these calculations was LiF (78 mole percent) - (HN)F 4 (22 mole percent) with a heavy nuclide composition of 3.86% {sup 233}U and 96.14% {sup 232}Th by mass. Before we can remove the fission product gases produced by this reactor configuration, we must first develop an appropriate storage mechanism. The gases could be stored in pressurized containers but then one must be concerned about bottle failure. Methods to trap noble gases in matrices are expensive and complex. Alternatively, there are direct storage/disposal options: direct injection into the Earth or injecting a grout-based product into the Earth. Advances in drilling technologies, hydro fracture technologies, and methods for the sequestration of carbon dioxide from fossil fuel plants are creating new options for disposal of fission gas wastes. In each option, lithostatic pressure, a kilometer or more underground, eliminates the pressure driving force for noble gas release and dissolves any untrapped gas in deep groundwater or into incorporated solid waste forms. The options, challenges, and potential for these methods to dispose of gaseous fission products are described. With this research, we hope to help both MSRs and other advanced reactors come one step closer to commercialization. (authors)

Messenger, S. J. [Massachusetts Inst. of Technology, 77 Massachusetts Ave., 54-1717, Cambridge, MA 02139 (United States); Forsberg, C. [Massachusetts Inst. of Technology, 77 Massachusetts Ave., 24-207, Cambridge, MA 02139 (United States); Massie, M. [Massachusetts Inst. of Technology, 77 Massachusetts Ave., NW12-230, Cambridge, MA 02139 (United States)

2012-07-01T23:59:59.000Z

269

Radiation safety assessment of a system of small reactors for distributed energy  

Science Journals Connector (OSTI)

......condition that the reactor was operated at full...Coolant water 55.0 Reactor vessel 30.0 Containment...at utilisation of nuclear energy other than electricity generation(3). Two PSRD units of 100 MW thermal reactor power are located......

N. Odano; T. Ishida

2005-12-20T23:59:59.000Z

270

XPS Investigations of Ruthenium Deposited onto Representative Inner Surfaces of Nuclear Reactor Containment Buildings  

E-Print Network (OSTI)

XPS Investigations of Ruthenium Deposited onto Representative Inner Surfaces of Nuclear Reactor in a nuclear power plant, interactions of gaseous RuO4 with reactor containment building surfaces (stainless, during nuclear reactor operation, the fission-product ruthenium will accumulate in the fuel. The quantity

Paris-Sud XI, Université de

271

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

E-Print Network (OSTI)

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

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

2001-10-23T23:59:59.000Z

272

Operations Information  

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

Standards BPA Operations Information (OPI) Transmission Services operates and plans for regional and national system needs. Transmission Services coordinates system operation and...

273

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

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

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

274

TYPICAL HOT WATER DRAW PATTERNS BASED ON FIELD DATA  

SciTech Connect

There is significant variation in hot water use and draw patterns among households. This report describes typical hot water use patterns in single-family residences in North America. We found that daily hot water use is highly variable both among residences and within the same residence. We compared the results of our analysis of the field data to the conditions and draw patterns established in the current U.S. Department of Energy (DOE) test procedure for residential water heaters. The results show a higher number of smaller draws at lower flow rates than used in the test procedure. The data from which the draw patterns were developed were obtained from 12 separate field studies. This report describes the ways in which we managed, cleaned, and analyzed the data and the results of our data analysis. After preparing the data, we used the complete data set to analyze inlet and outlet water temperatures. Then we divided the data into three clusters reflecting house configurations that demonstrated small, medium, or large median daily hot water use. We developed the three clusters partly to reflect efforts of the ASHRAE standard project committee (SPC) 118.2 to revise the test procedure for residential water heaters to incorporate a range of draw patterns. ASHRAE SPC 118.2 has identified the need to separately evaluate at least three, and perhaps as many as five, different water heater capacities. We analyzed the daily hot water use data within each cluster in terms of volume and number of hot water draws. The daily draw patterns in each cluster were characterized using distributions for volume of draws, duration of draws, time since previous draw, and flow rates.

Lutz, Jim; Melody, Moya

2012-11-08T23:59:59.000Z

275

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),

276

Investigation on tritium breeding capability of solid-liquid breeders in a (DT) fusion driven hybrid reactor  

Science Journals Connector (OSTI)

In design of deuterium-tritium (DT) fusion driven hybrid reactors, maintaining tritium self-sufficiency is very important ... to decrease operational costs. A (DT) fusion driven hybrid reactor must produce its ow...

M. Übeyli

2006-10-01T23:59:59.000Z

277

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

278

Novel Catalytic Membrane Reactors  

SciTech Connect

There are many industrial catalytic organic reversible reactions with amines or alcohols that have water as one of the products. Many of these reactions are homogeneously catalyzed. In all cases removal of water facilitates the reaction and produces more of the desired chemical product. By shifting the reaction to right we produce more chemical product with little or no additional capital investment. Many of these reactions can also relate to bioprocesses. Given the large number of water-organic compound separations achievable and the ability of the Compact Membrane Systems, Inc. (CMS) perfluoro membranes to withstand these harsh operating conditions, this is an ideal demonstration system for the water-of-reaction removal using a membrane reactor. Enhanced reaction synthesis is consistent with the DOE objective to lower the energy intensity of U.S. industry 25% by 2017 in accord with the Energy Policy Act of 2005 and to improve the United States manufacturing competitiveness. The objective of this program is to develop the platform technology for enhancing homogeneous catalytic chemical syntheses.

Stuart Nemser, PhD

2010-10-01T23:59:59.000Z

279

The ARIES tokamak reactor study  

SciTech Connect

The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

Not Available

1989-10-01T23:59:59.000Z

280

STATEMENT OF CONSIDERATIONS Advance Test Reactor Class Waiver  

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

Advance Test Reactor Class Waiver Advance Test Reactor Class Waiver W(C)-2008-004 The Advanced Test Reactor (A TR) is a pressurized water test reactor at the Idaho National Laboratory (INL) that operates at low pressure and temperature. The ATR was originally designed to study the effects of intense radiation on reactor material and fuels . It has a "Four Leaf Clover" design that allows a diverse array of testing locations. The unique design allows for different flux in various locations and specialized systems also allow for certain experiments to be run at their own temperature and pressure. The U.S. Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007. This designation will allow the ATR to

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


281

The High Flux Isotope Reactor at Oak Ridge National Laboratory  

NLE Websites -- All DOE Office Websites

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

282

Secretary Chu Statement on AP1000 Reactor Design Certification | Department  

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

Secretary Chu Statement on AP1000 Reactor Design Certification Secretary Chu Statement on AP1000 Reactor Design Certification Secretary Chu Statement on AP1000 Reactor Design Certification December 22, 2011 - 3:25pm Addthis Washington, D.C. - U.S. Energy Secretary Steven Chu issued the following statement today in support of the Nuclear Regulatory Commission's (NRC) decision to certify Westinghouse Electric's AP1000 nuclear reactor design, a significant step towards constructing a new generation of U.S. nuclear reactors. In February 2010, the Obama Administration announced the offer of a conditional commitment for a $8.33 billion loan guarantee for the construction and operation of two AP1000 reactors at Alvin W. Vogtle Electric Generation Plant in Burke, Georgia. "The Administration and the Energy Department are committed to restarting

283

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

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

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

284

Probabilistic risk assessment of N Reactor using NUREG-1150 methods  

SciTech Connect

A Level III probabilistic risk assessment (PRA) has been performed for N Reactor, a US Department of Energy (DOE) Category A production reactor. The main contractor is Westinghouse Hanford Company (Westinghouse Hanford). The PRA methodology developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL) in support of the NUREG-1150 (Reference 1) effort were used for this analysis. N Reactor is a graphite-moderated pressurized water reactor designed by General Electric. The dual-purpose 4000 MWt nuclear plant is located within the Hanford Site in the south-central part of the State of Washington. In addition to producing special materials for the DOE, N Reactor generates 860 MWe for the Washington Public Power Supply System. The reactor has been operated successfully and safely since 1963, and was put into standby status in 1988 due to the changing need in special nuclear material. 3 refs., 4 tabs.

Wang, O.S.; Baxter, J.T.; Coles, G.A.; Powers, T.B.; Zentner, M.D.

1989-11-01T23:59:59.000Z

285

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

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

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

286

Antineutrino monitoring for the Iranian heavy water reactor  

E-Print Network (OSTI)

In this note we discuss the potential application of antineutrino monitoring to the Iranian heavy water reactor at Arak, the IR-40, as a non-proliferation measure. We demonstrate that an above ground detector positioned right outside the IR-40 reactor building could meet and in some cases significantly exceed the verification goals identified by IAEA for plutonium production or diversion from declared inventories. In addition to monitoring the reactor during operation, observing antineutrino emissions from long-lived fission products could also allow monitoring the reactor when it is shutdown. Antineutrino monitoring could also be used to distinguish different levels of fuel enrichment. Most importantly, these capabilities would not require a complete reactor operational history and could provide a means to re-establish continuity of knowledge in safeguards conclusions should this become necessary.

Christensen, Eric; Jaffke, Patrick; Shea, Thomas

2014-01-01T23:59:59.000Z

287

AEC Pushes Fusion Reactors  

Science Journals Connector (OSTI)

AEC Pushes Fusion Reactors ... Project Sherwood, as the study program is called, began in 1951-52 soon after the first successful thermonuclear explosion in the Pacific. ...

1955-10-10T23:59:59.000Z

288

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

289

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

290

Hydrotreating operations discussed at refining meeting  

SciTech Connect

At the most recent National Petroleum Refiners Association question and answer session on refining and petrochemical technology, refiners and a panel of experts exchanged experiences on hydrotreater operations. Topics addressed included reactor pressurization, scale basket removal, and the use of antifoulants in effluent exchangers. This article presents comments from the panelists on the following questions. (1) What is the industry practice used to speed up the pressurization of 2.25 Cr/1 Mo reactors during start-up? Is there any relationship between reactor skin temperature and pressure used? (2) Has anyone removed scale baskets from a hydrotreating reactor and compared operations before and after? If so, were there any noticeable differences? Why? (3) What is the industry experience with the use of antifoulants for hydrocracking or hydrotreating reactor effluent exchangers?

NONE

1995-06-12T23:59:59.000Z

291

Plasma abatement of perfluorocompounds in inductively coupled plasma reactors  

E-Print Network (OSTI)

Plasma abatement of perfluorocompounds in inductively coupled plasma reactors Xudong ``Peter'' Xu PFCs , gases which have large global warming potentials, are widely used in plasma processing, the effluents from plasma tools using these gases typically have large mole fractions of PFCs. The use of plasma

Kushner, Mark

292

Process for operating equilibrium controlled reactions  

DOE Patents (OSTI)

A cyclic process for operating an equilibrium controlled reaction in a plurality of reactors containing an admixture of an adsorbent and a reaction catalyst suitable for performing the desired reaction which is operated in a predetermined timed sequence wherein the heating and cooling requirements in a moving reaction mass transfer zone within each reactor are provided by indirect heat exchange with a fluid capable of phase change at temperatures maintained in each reactor during sorpreaction, depressurization, purging and pressurization steps during each process cycle.

Nataraj, Shankar (Allentown, PA); Carvill, Brian Thomas (Orefield, PA); Hufton, Jeffrey Raymond (Fogelsville, PA); Mayorga, Steven Gerard (Allentown, PA); Gaffney, Thomas Richard (Allentown, PA); Brzozowski, Jeffrey Richard (Bethlehem, PA)

2001-01-01T23:59:59.000Z

293

FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL  

SciTech Connect

The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

2009-03-10T23:59:59.000Z

294

A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor...  

Office of Scientific and Technical Information (OSTI)

A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor Design and Feasibility Problem Re-direct Destination: Temp Data Fields Rosen, M. A.; Coburn, D. B.; Flynn, T....

295

Portfolio for fast reactor collaboration  

SciTech Connect

The development of the LMFBR type reactor in the United Kingdom is reviewed. Design characteristics of a commercial demonstration fast reactor are presented and compared with the Super Phenix reactor.

Rippon, S.

1981-12-01T23:59:59.000Z

296

Handbook of Reactor Physics  

Science Journals Connector (OSTI)

... THIS handbook is one volume in a series sponsored by the United States Atomic Energy Commission with ... data and reference information in the field of reactors. The volume is devoted to reactor physics and radiation shielding, the latter subject occupying approximately a quarter of the book.

PETER W. MUMMERY

1956-08-25T23:59:59.000Z

297

Fast reactor safety  

Science Journals Connector (OSTI)

... SIR, - In his article on fast reactor safety (26 July, page 270) Norman Dombey claims to introduce to non-specialists ... , page 270) Norman Dombey claims to introduce to non-specialists some features of fast reactors that are not available outside the technical literature. The non-specialist would do well ...

R.D. SMITH

1979-08-23T23:59:59.000Z

298

Instrumentation of Nuclear Reactors  

Science Journals Connector (OSTI)

... s Lecture Theatre on January 8, a symposium of papers on the instrumentation of nuclear reactors was organized, at which about five hundred members and visitors attended, including guests from ... the Institution, took the chair and introduced Sir John Cockcroft, whose lecture on "Nuclear Reactors and their Applications" provided a general background for the three specialized papers which followed. ...

1953-03-07T23:59:59.000Z

299

Nuclear Research Reactors  

Science Journals Connector (OSTI)

... their countries for the advent of nuclear power. A few countries had built large research reactors for the production of isotopes and to study the behaviour of nuclear fuel, but ... production of isotopes and to study the behaviour of nuclear fuel, but the small training reactor had not been developed. Since then, research ...

T. E. ALLIBONE

1963-07-20T23:59:59.000Z

300

Ames Laboratory Research Reactor Facility Ames, Iowa  

Office of Legacy Management (LM)

,, *' ; . Final Radiological Condition of the Ames Laboratory Research Reactor Facility Ames, Iowa _, . AGENCY: Office of Operational Safety, Department of Energy ' ACTION: Notice of Availability of Archival Information Package SUMMARY: The'Office of Operational Safety of the Department O i Energy (DOE) has reviewed documentation relating to the decontamination and decommissioning operations conducted at the Ames Laboratory Research Reactor Facility, Ames, Iowa and has prepared an archival informati0.n package to permanently document the results of the action and the site conditions and use restriction placed on the . site at the tim e of release. This review is based on post-decontamination survey data and other pertinent documentation referenced in and included in the archival package. The material and

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


301

TMI-2 reactor vessel head removal  

SciTech Connect

This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities.

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1984-12-01T23:59:59.000Z

302

TMI-2 reactor vessel head removal  

SciTech Connect

This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities.

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1985-09-01T23:59:59.000Z

303

Decommissioning of the Tokamak Fusion Test Reactor  

SciTech Connect

The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

2003-10-28T23:59:59.000Z

304

Modeling issues associated with production reactor safety assessment  

SciTech Connect

This paper describes several Probabilistic Safety Assessment (PSA) modeling issues that are related to the unique design and operation of the production reactors. The identification of initiating events and determination of a set of success criteria for the production reactors is of concern because of their unique design. The modeling of accident recovery must take into account the unique operation of these reactors. Finally, a more thorough search and evaluation of common-cause events is required to account for combinations of unique design features and operation that might otherwise not be included in the PSA. It is expected that most of these modeling issues also would be encountered when modeling some of the other more unique reactor and nonreactor facilities that are part of the DOE nuclear materials production complex. 9 refs., 2 figs.

Stack, D.W. (Los Alamos National Lab., NM (USA)); Thomas, W.R. (Science and Engineering Associates, Inc., Albuquerque, NM (USA))

1990-01-01T23:59:59.000Z

305

Reactor & Nuclear Systems Publications | ORNL  

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

Nuclear Science Home | Science & Discovery | Nuclear Science | Publications and Reports | Reactor and Nuclear Systems Publications SHARE Reactor and Nuclear Systems Publications...

306

The hybrid reactor project based on the straight field line mirror concept  

SciTech Connect

The straight field line mirror (SFLM) concept is aiming towards a steady-state compact fusion neutron source. Besides the possibility for steady state operation for a year or more, the geometry is chosen to avoid high loads on materials and plasma facing components. A comparatively small fusion hybrid device with 'semi-poor' plasma confinement (with a low fusion Q factor) may be developed for industrial transmutation and energy production from spent nuclear fuel. This opportunity arises from a large fission to fusion energy multiplication ratio, Q{sub r} = P{sub fis}/P{sub fus}>>1. The upper bound on Q{sub r} is primarily determined by geometry and reactor safety. For the SFLM, the upper bound is Q{sub r} Almost-Equal-To 150, corresponding to a neutron multiplicity of k{sub eff}=0.97. Power production in a mirror hybrid is predicted for a substantially lower electron temperature than the requirement T{sub e} Almost-Equal-To 10 keV for a fusion reactor. Power production in the SFLM seems possible with Q Almost-Equal-To 0.15, which is 10 times lower than typically anticipated for hybrids (and 100 times smaller than required for a fusion reactor). This relaxes plasma confinement demands, and broadens the range for use of plasmas with supra-thermal ions in hybrid reactors. The SFLM concept is based on a mirror machine stabilized by qudrupolar magnetic fields and large expander tanks beyond the confinement region. The purpose of the expander tanks is to distribute axial plasma loss flow over a sufficiently large area so that the receiving plates can withstand the heat. Plasma stability is not relying on a plasma flow into the expander regions. With a suppressed plasma flow into the expander tanks, a possibility arise for higher electron temperature. A brief presentation will be given on basic theory for the SFLM with plasma stability and electron temperature issues, RF heating computations with sloshing ion formation, neutron transport computations with reactor safety margins and material load estimates, magnetic coil designs as well as a discussion on the implications of the geometry for possible diagnostics. Reactor safety issues are addressed and a vertical orientation of the device could assist passive coolant circulation. Specific attention is put to a device with a 25 m long confinement region and 40 cm plasma radius in the mid-plane. In an optimal case (k{sub eff}= 0.97) with a fusion power of only 10 MW, such a device may be capable of producing a power of 1.5 GW{sub th}.

Agren, O.; Noack, K.; Moiseenko, V. E.; Hagnestal, A.; Kaellne, J.; Anglart, H. [Uppsala University, Angstroem Laboratory, Uppsala University, Box 534, SE-751 21 Uppsala (Sweden); Institute of Plasma Physics, National Science Center 'Kharkiv Institute of Physics and Technology', 61108 Kharkiv (Ukraine); Uppsala University, Angstroem Laboratory, Uppsala University, Box 534, SE-751 21 Uppsala (Sweden); Royal Institute of Technology, Nuclear Reactor Technology, SE 100 44 Stockholm (Sweden)

2012-06-19T23:59:59.000Z

307

Tritium issues in commercial pressurized water reactors  

SciTech Connect

Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

Jones, G. [Constellation Energy Group, R.E. Ginna Nuclear Power Plant, Ontario, NY (United States)

2008-07-15T23:59:59.000Z

308

Modifications to the NRAD Reactor, 1977 to present  

SciTech Connect

Argonne National Laboratory-West, operated by the University of Chicago, is located near Idaho Falls, ID, on the Idaho National Engineering laboratory Site. ANL-West performs work in support of the Liquid Metal Fast Breeder Reactor Program (LMFBR) sponsored by the United States Department of Energy. The NRAD reactor is located at the Argonne Site within the Hot Fuel Examination Facility/North, a large hot cell facility where both non-destructive and destructive examinations are performed on highly irradiated reactor fuels and materials in support of the LMFBR program. The NRAD facility utilizes a 250-kW TRIGA reactor and is completely dedicated to neutron radiography and the development of radiography techniques. Criticality was first achieved at the NRAD reactor in October of 1977. Since that time, a number of modifications have been implemented to improve operational efficiency and radiography production. This paper describes the modifications and changes that significantly improved operational efficiency and reliability of the reactor and the essential auxiliary reactor systems.

Weeks, A.A.; Pruett, D.P.; Heidel, C.C.

1986-01-01T23:59:59.000Z

309

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

310

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

311

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

312

Risk Management for Sodium Fast Reactors.  

SciTech Connect

Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.

Denman, Matthew R; Groth, Katrina; Cardoni, Jeffrey N; Wheeler, Timothy A.

2015-01-01T23:59:59.000Z

313

Small Modular Reactors (468th Brookhaven Lecture)  

SciTech Connect

With good reason, much more media attention has focused on nuclear power plants than solar farms, wind farms, or hydroelectric plants during the past month and a half. But as nations around the world demand more energy to power everything from cell phone batteries to drinking water pumps to foundries, nuclear plants are the only non-greenhouse-gas producing option that can be built to operate almost anywhere, and can continue to generate power during droughts, after the sun sets, and when winds die down. To supply this demand for power, designers around the world are competing to develop more affordable nuclear reactors of the future: small modular reactors. Brookhaven Lab is working with DOE to ensure that these reactors are designed to be safe for workers, members of surrounding communities, and the environment and to ensure that the radioactive materials and technology will only be used for peaceful purposes, not weapons. In his talk, Bari will discuss the advantages and challenges of small modular reactors and what drives both international and domestic interest in them. He will also explain how Brookhaven Lab and DOE are working to address the challenges and provide a framework for small modular reactors to be commercialized.

Bari, Robert

2011-04-20T23:59:59.000Z

314

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

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

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

315

Axial xenon stability considerations in VVER-1000 reactors  

SciTech Connect

Frequent problems experienced in VVER-1000 reactors with xenon oscillation control have been reported. Modern Pressurized Water Reactors (PWRs) are designed with operational strategies to help avoid such axial xenon oscillations. Uncontrolled xenon oscillations can cause operational problems, requiring frequent operator intervention and leading to power reductions (or plant trips) due to high core peaking factors, thereby reducing overall plant capacity factors. In the worst case, an uncontrolled xenon oscillation can lead to violations of safety requirements on pellet clad interaction (PCI), departure from nucleate boiling or fuel centerline melting if the plant does not have adequate safety protection systems that account for limiting axial power distributions.

Doshi, P.K.; Miller, R.W.

1993-12-31T23:59:59.000Z

316

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

317

Method of controlling crystallite size in nuclear-reactor fuels  

DOE Patents (OSTI)

Improved spherules for making enhanced forms of nuclear-reactor fuels are prepared by internal gelation procedures within a sol-gel operation and are accomplished by first boiling the concentrated HMTA-urea feed solution before engaging in the spherule-forming operation thereby effectively controlling crystallite size in the product spherules.

Lloyd, Milton H. (Oak Ridge, TN); Collins, Jack L. (Knoxville, TN); Shell, Sam E. (Oak Ridge, TN)

1985-01-01T23:59:59.000Z

318

The SUN Action database : collecting and analyzing typical actions for visual scene types  

E-Print Network (OSTI)

Recent work in human and machine vision has increasingly focused on the problem of scene recognition. Scene types are largely defined by the actions one might typically do there: an office is a place someone would typically ...

Olsson, Catherine Anne White

2013-01-01T23:59:59.000Z

319

DOE - Office of Legacy Management -- Reactor Site - Fort Belvoir - VA 0-02  

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

Reactor Site - Fort Belvoir - VA Reactor Site - Fort Belvoir - VA 0-02 FUSRAP Considered Sites Site: REACTOR SITE - FORT BELVOIR (VA.0-02 ) Eliminated from further consideration under FUSRAP - Referred to DOD Designated Name: Not Designated Alternate Name: None Location: Fort Belvoir , Virginia VA.0-02-1 Evaluation Year: 1987 VA.0-02-1 Site Operations: No evidence of AEC involvement with reactor operations. AEC conducted health and safety inspections of this site. Probably a licensed operation. VA.0-02-1 Site Disposition: Eliminated - Referred to DOD VA.0-02-1 Radioactive Materials Handled: Reactor fuel Primary Radioactive Materials Handled: Reactor Fuel Radiological Survey(s): Health and safety inspections VA.0-02-1 Site Status: Eliminated from further consideration under FUSRAP - Referred to DOD VA.0-02-1

320

Independent Confirmatory Survey Report for the University of Arizona Nuclear Reactor Laboratory, Tucson, Arizona  

SciTech Connect

The University of Arizona (University) research reactor is a TRIGA swimming pool type reactor designed by General Atomics and constructed at the University in 1958. The reactor first went into operation in December of 1958 under U.S. Nuclear Regulatory Commission (NRC) license R-52 until final shut down on May 18, 2010. Initial site characterization activities were conducted in February 2009 during ongoing reactor operations to assess the radiological status of the Nuclear Reactor Laboratory (NRL) excluding the reactor tank, associated components, and operating systems. Additional post-shutdown characterization activities were performed to complete characterization activities as well as verify assumptions made in the Decommissioning Plan (DP) that were based on a separate activation analysis (ESI 2009 and WMG 2009). Final status survey (FSS) activities began shortly after the issuance of the FSS plan in May 2011. The contractor completed measurement and sampling activities during the week of August 29, 2011.

Nick A. Altic

2011-11-11T23:59:59.000Z

Note: This page contains sample records for the topic "reactor operators typically" 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

Molten metal reactors  

DOE Patents (OSTI)

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

322

Light Water Reactors [Corrosion and Mechanics of Materials] - Nuclear  

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

Light Water Reactors Light Water Reactors Capabilities Materials Testing Environmentally Assisted Cracking (EAC) of Reactor Materials Corrosion Performance/Metal Dusting Overview Light Water Reactors Fatigue Testing of Carbon Steels and Low-Alloy Steels Environmentally Assisted Cracking of Ni-Base Alloys Irradiation-Induced Stress Corrosion Cracking of Austenitic Stainless Steels Steam Generator Tube Integrity Program Air Oxidation Kinetics for Zr-based Alloys Fossil Energy Fusion Energy Metal Dusting Publications List Irradiated Materials Steam Generator Tube Integrity Other Facilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Corrosion and Mechanics of Materials Light Water Reactors Bookmark and Share To continue safe operation of current LWRs, the aging degradation of the

323

Electric Power Produced from Nuclear Reactor | National Nuclear Security  

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

Electric Power Produced from Nuclear Reactor | National Nuclear Security Electric Power Produced from Nuclear Reactor | 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 > About Us > Our History > NNSA Timeline > Electric Power Produced from Nuclear Reactor Electric Power Produced from Nuclear Reactor December 20, 1951 Arco, ID Electric Power Produced from Nuclear Reactor

324

Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear  

National Nuclear Security Administration (NNSA)

Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear Maria Research Reactor loaded with LEU - Otwock, Poland | 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 > Media Room > Video Gallery > Maria Research Reactor loaded with LEU - ... Maria Research Reactor loaded with LEU - Otwock, Poland Maria Research Reactor loaded with LEU - Otwock, Poland

325

Operations Risk Management by Planning Optimally the Qualified ...  

E-Print Network (OSTI)

The back office will prepare the contracts, conduct all the exchange of information in due time and ... Internal risk is typically linked to operations (so controllable).

2007-07-31T23:59:59.000Z

326

F Reactor Inspection  

ScienceCinema (OSTI)

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

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-11-24T23:59:59.000Z

327

F Reactor Inspection  

SciTech Connect

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

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-10-29T23:59:59.000Z

328

Light Water Reactor Sustainability Program - Integrated Program Plan |  

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

Light Water Reactor Sustainability Program - Integrated Program Light Water Reactor Sustainability Program - Integrated Program Plan Light Water Reactor Sustainability Program - Integrated Program Plan The Light Water Reactor Sustainability (LWRS) Program is a research and development (R&D) program sponsored by the U. S. Department of Energy (DOE), performed in close collaboration and cooperation with related industry R&D programs. The LWRS Program provides technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants, utilizing the unique capabilities of the national laboratory system. Sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer than-initially-licensed lifetime. It has two facets

329

Advanced Neutron Source Reactor thermal analysis of fuel plate defects  

SciTech Connect

The Advanced Neutron Source Reactor (ANSR) is a research reactor designed to provide the highest continuous neutron beam intensity of any reactor in the world. The present technology for determining safe operations were developed for the High Flux Isotope Reactor (HFIR). These techniques are conservative and provide confidence in the safe operation of HFIR. However, the more intense requirements of ANSR necessitate the development of more accurate, but still conservative, techniques. This report details the development of a Local Analysis Technique (LAT) that provides an appropriate approach. Application of the LAT to two ANSR core designs are presented. New theories of the thermal and nuclear behavior of the U{sub 3}Si{sub 2} fuel are utilized. The implications of lower fuel enrichment and of modifying the inspection procedures are also discussed. Development of the computer codes that enable the automate execution of the LAT is included.

Giles, G.E.

1995-08-01T23:59:59.000Z

330

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

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

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

331

Kinetic model of the MTG process taking into account the catalyst deactivation. Reactor simulation  

Science Journals Connector (OSTI)

A kinetic model for the deactivation of catalyst (based on a HZSM5 zeolite) in the transformation of methanol into gasoline is proposed from results obtained in an isothermal fixed bed integral reactor. The kinetic model takes into account the effect of the composition of the lumps of oxygenates, light olefins and rest of products on the catalyst deactivation along the reactor. The model allows for simulating the integral reactor and for studying the influence of the operating conditions on selectivity towards different lumps in the MTG process. The resukts have been experimentally proven in an isothermal integral reactor and are in agreement with the results of coke deposition along the reactor.

A.G. Gayubo; P.L. Benito; A.T. Aguayo; M. Castilla; J. Bilbao

1996-01-01T23:59:59.000Z

332

High Flux Isotope Reactor power upgrade status  

SciTech Connect

A return to 100-MW operation is being planned for the High Flux Isotope Reactor (HFIR). Recent improvements in fuel element manufacturing procedures and inspection equipment will be exploited to reduce hot spot and hot streak factors sufficiently to permit the power upgrade without an increase in primary coolant pressure. Fresh fuel elements already fabricated for future use are being evaluated individually for power upgrade potential based on their measured coolant channel dimensions.

Rothrock, R.B.; Hale, R.E. [Oak Ridge National Lab., TN (United States); Cheverton, R.D. [Delta-21 Resources Inc., Oak Ridge, TN (United States)

1997-03-01T23:59:59.000Z

333

Hybrid energy systems (HESs) using small modular reactors (SMRs)  

SciTech Connect

Large-scale nuclear reactors are traditionally operated for a singular purpose: steady-state production of dispatchable baseload electricity that is distributed broadly on the electric grid. While this implementation is key to a sustainable, reliable energy grid, small modular reactors (SMRs) offer new opportunities for increased use of clean nuclear energy for both electric and thermal ap plications in more locations – while still accommodating the desire to support renewable production sources.

S. Bragg-Sitton

2014-10-01T23:59:59.000Z

334

REACTOR GROUT THERMAL PROPERTIES  

SciTech Connect

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

335

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

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

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

336

Monitoring the Thermal Power of Nuclear Reactors with a Prototype Cubic Meter Antineutrino Detector  

E-Print Network (OSTI)

In this paper, we estimate how quickly and how precisely a reactor's operational status and thermal power can be monitored over hour to month time scales, using the antineutrino rate as measured by a cubic meter scale detector. Our results are obtained from a detector we have deployed and operated at 25 meter standoff from a reactor core. This prototype can detect a prompt reactor shutdown within five hours, and monitor relative thermal power to three percent within seven days. Monitoring of short-term power changes in this way may be useful in the context of International Atomic Energy Agency's (IAEA) Reactor Safeguards Regime, or other cooperative monitoring regimes.

A. Bernstein; N. S. Bowden; A. Misner; T. Palmer

2008-04-30T23:59:59.000Z

337

Reactor User Interface Technology Development Roadmaps for a High Temperature Gas-Cooled Reactor Outlet Temperature of 750 degrees C  

SciTech Connect

This report evaluates the technology readiness of the interface components that are required to transfer high-temperature heat from a High Temperature Gas-Cooled Reactor (HTGR) to selected industrial applications. This report assumes that the HTGR operates at a reactor outlet temperature of 750°C and provides electricity and/or process heat at 700°C to conventional process applications, including the production of hydrogen.

Ian Mckirdy

2010-12-01T23:59:59.000Z

338

Export possibilities for small nuclear reactors  

SciTech Connect

The worldwide deployment of peaceful nuclear technology is predicated on conformance with the Nuclear Non-Proliferation Treaty of 1972. Under this international treaty, countries have traded away pursuit of nuclear weapons in exchange for access to commercial nuclear technology that could help them grow economically. Realistically, however, most nuclear technology has been beyond the capacity of the NPT developing countries to afford. Even if the capital cost of the plant is managed, the costs of the infrastructure and the operational complexity of most nuclear technology have taken it out of the hands of the nations who need it the most. Now, a new class of small sodium cooled reactors has been specifically designed to meet the electrical power, water, hydrogen and heat needs of small and remote users. These reactors feature small size, long refueling interval, no onsite fuel storage, and simplified operations. Sized in the 10 MW(e) to 50 MW(e) range these reactors are modularized for factory production and for rapid site assembly. The fuel would be <20% U-235 uranium fuel with a 30-year core life. This new reactor type more appropriately fills the needs of countries for lower power distributed systems that can fill the gap between large developed infrastructure and primitive distributed energy systems. Looking at UN Resolution 1540 and the impact of other agreements, there is a need to address the issues of nuclear security, fuel, waste, and economic/legal/political-stakeholder concerns. This paper describes the design features of this new reactor type that specifically address these issues in a manner that increases the availability of commercial nuclear technology to the developing nations of the world. (authors)

Campagna, M.S.; Hess, C.; Moor, P. [Burns and Roe Enterprises, Inc., Oradell, NJ (United States); Sawruk, W. [ABSG Consulting, Inc., Shillington, PA (United States)

2007-07-01T23:59:59.000Z

339

The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow  

SciTech Connect

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

2008-07-15T23:59:59.000Z

340

Proliferation resistance of small modular reactors fuels  

SciTech Connect

In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. In the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.

Polidoro, F.; Parozzi, F. [RSE - Ricerca sul Sistema Energetico,Via Rubattino 54, 20134, Milano (Italy); Fassnacht, F.; Kuett, M.; Englert, M. [IANUS, Darmstadt University of Technology, Alexanderstr. 35, D-64283 Darmstadt (Germany)

2013-07-01T23:59:59.000Z

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


341

New AB-Thermonuclear Reactor for Aerospace  

E-Print Network (OSTI)

There are two main methods of nulcear fusion: inertial confinement fusion (ICF) and magnetic confinement fusion (MCF). Existing thermonuclear reactors are very complex, expensive, large, and heavy. They cannot achieve the Lawson creterion. The author offers an innovation. ICF has on the inside surface of the shell-shaped combustion chamber a covering of small Prism Reflectors (PR) and plasma reflector. These prism reflectors have a noteworthy advantage, in comparison with conventional mirror and especially with conventional shell: they multi-reflect the heat and laser radiation exactly back into collision with the fuel target capsule (pellet). The plasma reflector reflects the Bremsstrahlung radiation. The offered innovation decreases radiation losses, creates significant radiation pressure and increases the reaction time. The Lawson criterion increases by hundreds of times. The size, cost, and weight of a typical installation will decrease by tens of times. The author is researching the efficiency of these innovations. Keywords: Thermonuclear reactor, Multi-reflex AB-thermonuclear reactor, aerospace thermonuclear engine. This work is presented as paper AIAA-2006-7225 to Space-2006 Conference, 19-21 September, 2006, San Jose, CA, USA.

Alexander Bolonkin

2007-06-14T23:59:59.000Z

342

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

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

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

343

Design and Operational Considerations of Educational Satellite Systems  

Science Journals Connector (OSTI)

...The implications of hybrid video systems are examined, incorporating fixed...critical aspect of the educational system operation is likely to be the production...operational problems of a typical system are explored, emphasizing the...

1975-01-01T23:59:59.000Z

344

Thermionic Reactor Design Studies  

SciTech Connect

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

Schock, Alfred

1994-08-01T23:59:59.000Z

345

Fossil-fuel processing technical/professional services: comparison of Fischer-Tropsch reactor systems. Phase I, final report  

SciTech Connect

The Fischer-Tropsch reaction was commercialized in Germany and used to produce military fuels in fixed bed reactors. It was recognized from the start that this reactor system had severe operating and yield limitations and alternative reactor systems were sought. In 1955 the Sasol I complex, using an entrained bed (Synthol) reactor system, was started up in South Africa. Although this reactor was a definite improvement and is still operating, the literature is filled with proponents of other reactor systems, each claiming its own advantages. This report provides a summary of the results of a study to compare the development potential of three of these reactor systems with the commercially operating Synthol-entrained bed reactor system. The commercial Synthol reactor is used as a benchmark against which the development potential of the other three reactors can be compared. Most of the information on which this study is based was supplied by the M.W. Kellogg Co. No information beyond that in the literature on the operation of the Synthol reactor system was available for consideration in preparing this study, nor were any details of the changes made to the original Synthol system to overcome the operating problems reported in the literature. Because of conflicting claims and results found in the literature, it was decided to concentrate a large part of this study on a kinetic analysis of the reactor systems, in order to provide a theoretical analysis of intrinsic strengths and weaknesses of the reactors unclouded by different catalysts, operating conditions and feed compositions. The remainder of the study considers the physical attributes of the four reactor systems and compares their respective investment costs, yields, catalyst requirements and thermal efficiencies from simplified conceptual designs.

Thompson, G.J.; Riekena, M.L.; Vickers, A.G.

1981-09-01T23:59:59.000Z

346

OPERATIONS (OPS)  

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

OPS) OPS) OBJECTIVE OPS.1 The formality and discipline of operations is adequate to conduct work safely and programs are in place to maintain this formality and discipline. (CR 13) Scope: The Conduct of Operations Program was evaluated during the recent KE Basin FTS ORR and was found to be adequately implemented. Based on this result and the subsequent program enhancements, the scope of the review is to be limited to the SWS operating and maintenance evolutions. Criteria * Programmatic elements of conduct of operations are in place for SWS operations. (DOE Order 5480.19) * The SWS operations personnel adequately demonstrate the principles of conduct of operations requirements during the shift performance period. (DOE Order 5480.19)

347

Using the three-way catalyst monolith reactor for reducing exhaust emissions  

Science Journals Connector (OSTI)

The monolith reactor was developed for the cleaning of exhaust gases from combustion processes both in cars and large power plants. Nowadays monolith reactors are increasingly being used developed evaluated in automotive and stationary emission control reactors such as power plants and new reactor applications such as chemical and refining processes catalytic combustion ozone abatement and others. Monolith catalysts mainstays in gas-phase automotive and environmental process applications have found new potential in replacing three-phase slurry reactors for the production of specialty chemicals especially when their advantages are fully utilized in recirculation loop approaches. This paper gives a general overview about monolith reactors’ benefits fabrication characteristics and typical use in automotive industry. Several commercial product applications and new developments for use of monolith reactors in automotive stationary and chemical industry have been discussed. Different types of monolith reactor systems manufacturing modeling and application areas are specified with their advantages and disadvantages. Some experimental studies have been attached to compare monolith reactor types with conventional reactors.

Burak Gokalp

2012-01-01T23:59:59.000Z

348

TFTR tritium operations lessons learned  

SciTech Connect

The Tokamak Fusion Test Reactor which is the progenitor for full D-T operating tokamaks has successfully processed > 81 grams of tritium in a safe and efficient fashion. Many of the fundamental operational techniques associated with the safe movement of tritium through the TFTR facility were developed over the course of many years of DOE tritium facilities (LANL, LLNL, SRS, Mound). In the mid 1980`s The Tritium Systems Test Assembly (TSTA) at LANL began reporting operational techniques for the safe handling of tritium, and became a major conduit for the transfer of safe tritium handling technology from DOE weapons laboratories to non-weapon facilities. TFTR has built on many of the TSTA operational techniques and has had the opportunity of performing and enhancing these techniques at America`s first operational D-T fusion reactor. This paper will discuss negative pressure employing `elephant trunks` in the control and mitigation of tritium contamination at the TFTR facility, and the interaction between contaminated line operations and {Delta} pressure control. In addition the strategy employed in managing the movement of tritium through TFTR while maintaining an active tritium inventory of < 50,000 Ci will be discussed. 5 refs.

Gentile, C.A.; Raftopoulos, S.; LaMarche, P. [Princeton Plasma Physics Lab., NJ (United States)] [and others

1996-12-31T23:59:59.000Z

349

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

350

Refueling Liquid-Salt-Cooled Very High-Temperature Reactors  

SciTech Connect

The liquid-salt-cooled very high-temperature reactor (LS-VHTR), also called the Advanced High-Temperature Reactor (AHTR), is a new reactor concept that combines in a novel way four established technologies: (1) coated-particle graphite-matrix nuclear fuels, (2) Brayton power cycles, (3) passive safety systems and plant designs previously developed for liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants. Depending upon goals, the peak coolant operating temperatures are between 700 and 1000 deg. C, with reactor outputs between 2400 and 4000 MW(t). Several fluoride salt coolants that are being evaluated have melting points between 350 and 500 deg. C, values that imply minimum refueling temperatures between 400 and 550 deg. C. At operating conditions, the liquid salts are transparent and have physical properties similar to those of water. A series of refueling studies have been initiated to (1) confirm the viability of refueling, (2) define methods for safe rapid refueling, and (3) aid the selection of the preferred AHTR design. Three reactor cores with different fuel element designs (prismatic, pebble bed, and pin-type fuel assembly) are being evaluated. Each is a liquid-salt-cooled variant of a graphite-moderated high-temperature reactor. The refueling studies examined applicable refueling experience from high-temperature reactors (similar fuel element designs) and sodium-cooled fast reactors (similar plant design with liquid coolant, high temperatures, and low pressures). The findings indicate that refueling is viable, and several approaches have been identified. The study results are described in this paper. (authors)

Forsberg, Charles W. [Oak Ridge National Laboratory, P.O. Box 2008 Oak Ridge, TN 37831 (United States); Peterson, Per F. [Nuclear Engineering Department, University of California at Berkeley, 6124a Etcheverry Hall, Berkeley, CA 94720 (United States); Cahalan, James E. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Enneking, Jeffrey A. [Areva NP (United States); Phil MacDonald [Consultant, Cedar Hill, TX (United States)

2006-07-01T23:59:59.000Z

351

Radiation effects on reactor pressure vessel supports  

SciTech Connect

The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue.

Johnson, R.E. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology; Lipinski, R.E. [Idaho National Engineering Lab., Rockville, MD (United States)

1996-05-01T23:59:59.000Z

352

Development of pyro-processing technology for thorium-fuelled molten salt reactor  

SciTech Connect

The Molten Salt Reactor (MSR) is classified as the non-classical nuclear reactor type based on the specific features coming out from the use of liquid fuel circulating in the MSR primary circuit. Other uniqueness of the reactor type is based on the fact that the primary circuit of the reactor is directly connected with the on-line reprocessing technology, necessary for keeping the reactor in operation for a long run. MSR is the only reactor system, which can be effectively operated within the {sup 232}Th- {sup 233}U fuel cycle as thorium breeder with the breeding factor significantly higher than one. The fuel cycle technologies proposed as ford the fresh thorium fuel processing as for the primary circuit fuel reprocessing are pyrochemical and mainly fluoride. Although these pyrochemical processes were never previously fully verified, the present-day development anticipates an assumption for the successful future deployment of the thorium-fuelled MSR technology. (authors)

Uhlir, J.; Straka, M.; Szatmary, L. [Nuclear Research Inst. ReZ Plc, ReZ 130, Husinec - CZ-250 68 (Czech Republic)

2012-07-01T23:59:59.000Z

353

An improved weighted average reactor temperature estimation for simulation of adiabatic industrial hydrotreaters  

Science Journals Connector (OSTI)

A study on the improvement of the representative operating temperature from the temperature profile of an industrial adiabatic reactor is presented. This temperature is used to simulate the reactor performance by small scale laboratory isothermal reactors. An improved methodology for the estimation of a Weighted Average Bed Temperature (WABT) was elaborated to simulate an industrial multi-bed HDS reactor. The improved WABT, so called Weighted Average Reactor Temperature (WART), was compared with the most usually used WABT in a wide range of operational conditions as well as of kinetic parameters. In case of a multi-bed industrial hydrotreater, where quench zones are located between the beds and the H2 flow rate, which enters each bed, is different, the optimal gas to oil ratio was estimated for the laboratory-scale reactor.

G.D. Stefanidis; G.D. Bellos; N.G. Papayannakos

2005-01-01T23:59:59.000Z

354

Structural materials for fusion reactors  

Science Journals Connector (OSTI)

Fusion Reactors will require specially engineered structural materials, which ... on safety considerations. The fundamental differences between fusion and other nuclear reactors arise due to the 14MeV neutronics ...

P. M. Raole; S. P. Deshpande

2009-04-01T23:59:59.000Z

355

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

SciTech Connect

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

356

Reactor Materials | Department of Energy  

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

Benefits Crosscutting Technology Development Reactor Materials Advanced Sensors and Instrumentation Proliferation and Terrorism Risk Assessment Advanced Methods for Manufacturing...

357

Drought remedial measures through resistivity investigations in a typical crystalline region  

Science Journals Connector (OSTI)

Systematic geoelectrical investigations were carried out in a typical drought ... of Andhra Pradesh, India, for evolving drought remedial strategies. Depth to basement maps, geoelectrical...

B. H. Briz-Kishore

358

Five Years of Building the Next Generation of Reactors | Department of  

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

Five Years of Building the Next Generation of Reactors Five Years of Building the Next Generation of Reactors Five Years of Building the Next Generation of Reactors August 15, 2012 - 5:17pm Addthis Simulated three-dimensional fission power distribution of a single 17x17 rod PWR fuel assembly. | Photo courtesy of the Consortium for Advanced Simulation of Light Water Reactors (CASL). Simulated three-dimensional fission power distribution of a single 17x17 rod PWR fuel assembly. | Photo courtesy of the Consortium for Advanced Simulation of Light Water Reactors (CASL). Doug Kothe Director, Consortium for Advanced Simulation of Light Water Reactors What are the key facts? CASL has the virtual capability to look closely at reactor core models. These models operate with 193 fuel assemblies, nearly 51,000 fuel rods, and about 18 million fuel pellets.

359

Integrated intelligent systems in advanced reactor control rooms  

SciTech Connect

An intelligent, reactor control room, information system is designed to be an integral part of an advanced control room and will assist the reactor operator's decision making process by continuously monitoring the current plant state and providing recommended operator actions to improve that state. This intelligent system is an integral part of, as well as an extension to, the plant protection and control systems. This paper describes the interaction of several functional components (intelligent information data display, technical specifications monitoring, and dynamic procedures) of the overall system and the artificial intelligence laboratory environment assembled for testing the prototype. 10 refs., 5 figs.

Beckmeyer, R.R.

1989-01-01T23:59:59.000Z

360

Transport reactor development status  

SciTech Connect

This project is part of METC`s Power Systems Development Facility (PSDF) located at Wilsonville, Alabama. The primary objective of the Advanced Gasifier module is to produce vitiated gases for intermediate-term testing of Particulate Control Devices (PCDs). The Transport reactor potentially allows particle size distribution, solids loading, and particulate characteristics in the off-gas stream to be varied in a number of ways. Particulates in the hot gases from the Transport reactor will be removed in the PCDs. Two PCDs will be initially installed in the module; one a ceramic candle filter, the other a granular bed filter. After testing of the initial PCDs they will be removed and replaced with PCDs supplied by other vendors. A secondary objective is to verify the performance of a Transport reactor for use in advanced Integrated Gasification Combined Cycle (IGCC), Integrated Gasification Fuel Cell (IG-FC), and Pressurized Combustion Combined Cycle (PCCC) power generation units. This paper discusses the development of the Transport reactor design from bench-scale testing through pilot-scale testing to design of the Process Development Unit (PDU-scale) facility at Wilsonville.

Rush, R.E.; Fankhanel, M.O.; Campbell, W.M.

1994-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor operators typically" 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

Thermal Reactor Safety  

SciTech Connect

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

362

NETL - Chemical Looping Reactor  

ScienceCinema (OSTI)

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

None

2014-06-26T23:59:59.000Z

363

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

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

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

364

The development and demonstration of a thermal neutron radiography facility utilizing the TAMU NSC TRIGA reactor  

E-Print Network (OSTI)

of the beam port door and exposing detector at low reactor power. In order to permit operation 26 continously with minimal disruption of normal reactor operations and minimal radiation exposure to operating personnel, an object- detector holding apparatus... exposure except from the gamma rays in the beam. Righ quality radiographs were 29 obtained using Type R film with exposures of 20-30 minutes at a reactor power of 1 MW. Type M2/RP and Type AA films were used when speed was desired. Type AA...

Lorenz, Robert Wayne

2012-06-07T23:59:59.000Z

365

Programmable AC power supply for simulating power transient expected in fusion reactor  

SciTech Connect

This paper focus on control engineering of the programmable AC power source which has capability to simulate power transient expected in fusion reactor. To generate the programmable power source, AC-AC power electronics converter is adopted to control the power of a set of heaters to represent the transient phenomena of heat exchangers or heat sources of a fusion reactor. The International Thermonuclear Experimental Reactor (ITER) plasma operation scenario is used as the basic reference for producing this transient power source. (authors)

Halimi, B. [Seoul National Univ., Seoul 151-744 (Korea, Republic of); Suh, K. Y. [Seoul National Univ., Seoul 151-744 (Korea, Republic of); PHILOSOPHIA, 1 Gwanak Ro, Gwanak Gu, Seoul 151-744 (Korea, Republic of)

2012-07-01T23:59:59.000Z

366

SPEAR Operations  

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

Interface 1113 N. Kurita J. Langton Vacuum TSP's 1120 J. Corbett A. Terebilo MATLAB Applications - Basics 1121 F. Rafael Booster Kicker Upgrade, Operation Manual 1121...

367

operations center  

National Nuclear Security Administration (NNSA)

1%2A en Operations Center http:nnsa.energy.govaboutusourprogramsemergencyoperationscounterterrorismoperationscenter

...

368

Prospects for the development of advanced reactors  

SciTech Connect

Energy supply is an important prerequisite for further socio-economic development, especially in developing countries where the per capita energy use is only a very small fraction of that in industrialized countries. Nuclear energy is an essentially unlimited energy resource with the potential to provide this energy in the form of electricity, district heat and process heat under environmentally acceptable conditions. However, this potential will be realized only if nuclear power plants can meet the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide a tremendous amount of experience has been accumulated during development, licensing, construction and operation of nuclear power reactors. The experience forms a sound basis for further improvements. Nuclear programmes in many countries are addressing the development of advanced reactors which are intended to have better economics, higher reliability and improved safety in order to overcome the current concerns of nuclear power. Advanced reactors now being developed could help to meet the demand for new plants in developed and developing countries, not only for electricity generation, but also for district heating, desalination and for process heat. The IAEA, as the only global international governmental organization dealing with nuclear power, promotes international information exchange and international co-operation between all countries with their own advanced nuclear power programmes and offers assistance to countries with an interest in exploratory or research programmes.

Semenov, B.A.; Kupitz, J.; Cleveland, J. [International Atomic Energy Agency Vienna (Austria). Dept. of Nuclear Energy and Safety

1992-12-31T23:59:59.000Z

369

Operational Experience in Nuclear Power Stations [and Discussion  

Science Journals Connector (OSTI)

...Operational Experience in Nuclear Power Stations...self-sustaining nuclear reaction to the present...time large-scale generation of electrical power from nuclear energy has become...the C.E.G.B. reactors have been in service...

1974-01-01T23:59:59.000Z

370

Methods for quantifying uncertainty in fast reactor analyses.  

SciTech Connect

Liquid-metal-cooled fast reactors in the form of sodium-cooled fast reactors have been successfully built and tested in the U.S. and throughout the world. However, no fast reactor has operated in the U.S. for nearly fourteen years. More importantly, the U.S. has not constructed a fast reactor in nearly 30 years. In addition to reestablishing the necessary industrial infrastructure, the development, testing, and licensing of a new, advanced fast reactor concept will likely require a significant base technology program that will rely more heavily on modeling and simulation than has been done in the past. The ability to quantify uncertainty in modeling and simulations will be an important part of any experimental program and can provide added confidence that established design limits and safety margins are appropriate. In addition, there is an increasing demand from the nuclear industry for best-estimate analysis methods to provide confidence bounds along with their results. The ability to quantify uncertainty will be an important component of modeling that is used to support design, testing, and experimental programs. Three avenues of UQ investigation are proposed. Two relatively new approaches are described which can be directly coupled to simulation codes currently being developed under the Advanced Simulation and Modeling program within the Reactor Campaign. A third approach, based on robust Monte Carlo methods, can be used in conjunction with existing reactor analysis codes as a means of verification and validation of the more detailed approaches.

Fanning, T. H.; Fischer, P. F.

2008-04-07T23:59:59.000Z

371

Gas-cooled nuclear reactor  

DOE Patents (OSTI)

A gas-cooled nuclear reactor includes a central core located in the lower portion of a prestressed concrete reactor vessel. Primary coolant gas flows upward through the core and into four overlying heat-exchangers wherein stream is generated. During normal operation, the return flow of coolant is between the core and the vessel sidewall to a pair of motor-driven circulators located at about the bottom of the concrete pressure vessel. The circulators repressurize the gas coolant and return it back to the core through passageways in the underlying core structure. If during emergency conditions the primary circulators are no longer functioning, the decay heat is effectively removed from the core by means of natural convection circulation. The hot gas rising through the core exits the top of the shroud of the heat-exchangers and flows radially outward to the sidewall of the concrete pressure vessel. A metal liner covers the entire inside concrete surfaces of the concrete pressure vessel, and cooling tubes are welded to the exterior or concrete side of the metal liner. The gas coolant is in direct contact with the interior surface of the metal liner and transfers its heat through the metal liner to the liquid coolant flowing through the cooling tubes. The cooler gas is more dense and creates a downward convection flow in the region between the core and the sidewall until it reaches the bottom of the concrete pressure vessel when it flows radially inward and up into the core for another pass. Water is forced to flow through the cooling tubes to absorb heat from the core at a sufficient rate to remove enough of the decay heat created in the core to prevent overheating of the core or the vessel.

Peinado, Charles O. (La Jolla, CA); Koutz, Stanley L. (San Diego, CA)

1985-01-01T23:59:59.000Z

372

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

373

The BGU/CERN solar hydrothermal reactor  

E-Print Network (OSTI)

We describe a novel solar hydrothermal reactor (SHR) under development by Ben Gurion University (BGU) and the European Organization for Nuclear Research CERN. We describe in broad terms the several novel aspects of the device and, by extension, of the niche it occupies: in particular, enabling direct off-grid conversion of a range of organic feedstocks to sterile useable (solid, liquid) fuels, nutrients, products using only solar energy and water. We then provide a brief description of the high temperature high efficiency panels that provide process heat to the hydrothermal reactor, and review the basics of hydrothermal processes and conversion taking place in this. We conclude with a description of a simulation of the pilot system that will begin operation later this year.

Bertolucci, Sergio; Caspers, Fritz; Garb, Yaakov; Gross, Amit; Pauletta, Stefano

2014-01-01T23:59:59.000Z

374

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

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

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

375

Magnetically Controlled Reactor Shrinks Power Quality Costs and Power  

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

Magnetically Controlled Reactor Shrinks Power Quality Costs and Power Magnetically Controlled Reactor Shrinks Power Quality Costs and Power Losses Speaker(s): Mark D. Galperin Date: December 18, 2000 - 12:00pm Location: Bldg. 90 Seminar Host/Point of Contact: Diana Morris In a new, magnetically controlled reactor (MCR), in which DC pulsing through a special winding controls inductive susceptance, high saturation of the magnetic circuit steel with optimal magnetic and electrical circuit parameters ensures less than 2-3% main harmonic distortion even without special filters. Transformer-like construction ensures reliable operation. MCR's increase power quality through automatic voltage regulation, reduced fluctuation, and smoothing of reactive power surges at 1/2 the cost of thyristor-controlled reactors (TCR's). Damping of voltage-oscillation

376

Cost-Shared Development of Innovative Small Modular Reactor Designs |  

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

Cost-Shared Development of Innovative Small Modular Reactor Designs Cost-Shared Development of Innovative Small Modular Reactor Designs Cost-Shared Development of Innovative Small Modular Reactor Designs The Small Modular Reactor (SMR) Licensing Technical Support (LTS) program, sponsored by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), through this Funding Opportunity Announcement (FOA) seeks to facilitate the development of innovative SMR designs that have the potential to address the nation's economic, environmental and energy security goals. Specifically, the Department is soliciting applications for SMR designs that offer unique and innovative solutions for achieving the objectives of enhanced safety, operations, and performance relative to currently certified designs. This FOA focuses on design development and

377

Vertical Pretreatment Reactor System (Poster), NREL (National Renewable Energy Laboratory)  

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

Vertical Pretreatment Reactor System Vertical Pretreatment Reactor System Two-vessel system for primary and secondary pretreatment at diff erent temperatures * Biomass is heated by steam injection to temperatures of 120°C to 210°C in the pressurized mixing tube * Preheated, premixed biomass is retained for specified residence time in vertical holding vessel; material continuously moves by gravity from top to bottom of reactor in plug-fl ow fashion * Residence time is adjusted by changing amount of material held in vertical vessel relative to continuous fl ow of material entering and exiting vessel * Optional additional reactor vessel allows for secondary pretreatment at lower temperatures-120°C to 180°C-with potential to add other chemical catalysts * First vessel can operate at residence

378

DOE - Office of Legacy Management -- Westinghouse Advanced Reactors Div  

Office of Legacy Management (LM)

Advanced Reactors Div Advanced Reactors Div Plutonium and Advanced Fuel Labs - PA 10 FUSRAP Considered Sites Site: WESTINGHOUSE ADVANCED REACTORS DIV., PLUTONIUM FUEL LABORATORIES, AND THE ADVANCED FUEL LAB (PA.10 ) Eliminated from further consideration under FUSRAP Designated Name: Not Designated Alternate Name: None Location: Cheswick , Pennsylvania PA.10-1 Evaluation Year: Circa 1987 PA.10-1 PA.10-4 Site Operations: 1960s and 1970s - Produced light water and fast breeder reactor fuels on a development and pilot plant scale. Closed in 1979. PA.10-2 PA.10-3 Site Disposition: Eliminated - Decommissioned and decontaminated under another Federal program. Release condition confirmed by radiological surveys. PA.10-1 PA.10-2 PA.10-3 PA.10-4 PA.10-5 Radioactive Materials Handled: Yes

379

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

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

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

380

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

SciTech Connect

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

Rothrock, R.B.

1991-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor operators typically" from the National Library of EnergyBeta (NLEBeta).
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381

Neutron damage reduction in a traveling wave reactor  

SciTech Connect

Traveling wave reactors are envisioned to run on depleted or natural uranium with no need for enrichment or reprocessing, and in a manner which requires little to no operator intervention. If feasible, this type of reactor has significant advantages over conventional nuclear power systems. However, a practical implementation of this concept is challenging as neutron irradiation levels many times greater than those in conventional reactors appear to be required for a fission wave to propagate. Radiation damage to the fuel and cladding materials presents a significant obstacle to a practical design. One possibility for reducing damage is to soften the neutron energy spectrum. Here we show that using a uranium oxide fuel form will allow a shift in the neutron spectrum that can result in at least a three fold decrease in dpa levels for fuel cladding and structural steels within the reactor compared with the dpa levels expected when using a uranium metal fuel. (authors)

Osborne, A. G.; Deinert, M. R. [Dept. of Mechanical Engineering, Univ. of Texas at Austin, Austin, TX (United States)

2012-07-01T23:59:59.000Z

382

Dual annular rotating "windowed" nuclear reflector reactor control system  

DOE Patents (OSTI)

A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core.

Jacox, Michael G. (Idaho Falls, ID); Drexler, Robert L. (Idaho Falls, ID); Hunt, Robert N. M. (Idaho Falls, ID); Lake, James A. (Idaho Falls, ID)

1994-01-01T23:59:59.000Z

383

Cost-Shared Development of Innovative Small Modular Reactor Designs |  

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

Cost-Shared Development of Innovative Small Modular Reactor Designs Cost-Shared Development of Innovative Small Modular Reactor Designs Cost-Shared Development of Innovative Small Modular Reactor Designs The Small Modular Reactor (SMR) Licensing Technical Support (LTS) program, sponsored by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), through this Funding Opportunity Announcement (FOA) seeks to facilitate the development of innovative SMR designs that have the potential to address the nation's economic, environmental and energy security goals. Specifically, the Department is soliciting applications for SMR designs that offer unique and innovative solutions for achieving the objectives of enhanced safety, operations, and performance relative to currently certified designs. This FOA focuses on design development and

384

The Reactor engineering of the MITR-II : construction and startup  

E-Print Network (OSTI)

The heavy water moderated and cooled research reactor, MITR-I, has been replaced with a light water cooled, heavy water reflected reactor called the MITR-II. The MITR-II is designed to operate at 5 thermal megawatts. The ...

Allen, G. C.

1976-01-01T23:59:59.000Z

385

Performance Analysis of the Reactor Pattern in Network Services Swapna Gokhale Aniruddha Gokhale Jeff Gray  

E-Print Network (OSTI)

Performance Analysis of the Reactor Pattern in Network Services Swapna Gokhale Aniruddha Gokhale into a reusable pattern, such as the Reactor pattern. In order to enable performance analysis of event by software applications places a high premium on the reliable and efficient operation of these applications

Gray, Jeffrey G.

386

A model for waterside oxidation of zircaloy fuel cladding in pressurized water reactors  

SciTech Connect

A model is developed to simulate the oxidation of zircaloy fuel rod cladding exposed to pressurized water reactor operating conditions. The model is used to predict the oxidation rate for both ex- and in-reactor conditions in terms of the weight gain and oxide thickness. Comparisons of the model predictions with experimental data show very good agreement.

Amarshad, A.I.A. [Institute of Atomic Energy Research, Riyadh (Saudi Arabia); Klein, A.C. [Oregon State Univ., Corvallis, OR (United States)

1992-12-31T23:59:59.000Z

387

Safety of CANDU reactors utilizing plutonium-enriched mixed-oxide fuel  

SciTech Connect

Substantial quantities of plutonium have become available as a result of nuclear arms reduction agreements. Irradiation of plutonium enriched fuel in Canadian deuterium uranium (CANDU) heavy water moderated and cooled reactors, of which there are 22 in operation in Canada, has been evaluated as a means of managing it. This paper summarizes the results of a study of reactor safety.

Chan, P.; Feinroth, H.; Luxat, J.; Pendergast, D.

1994-12-31T23:59:59.000Z

388

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

389

Fast spectrum space reactor sizing code for calandria-type cores (CORSCO Code). [Li  

SciTech Connect

The CORSCO code rapidly sizes reactor cores that have calandria-type geometry. The fuel configuration modeled is a large ceramic zone that contains numerous small cylindrical coolant channels spaced apart with a triangular pitch. A minimum reactor weight is obtained for a fixed set of constraints (peak fuel temperature, peak coolant velocity, etc.) by obtaining a unique solution to a set of five thermal/hydraulic equations, as well as a required excess reactivity which is specified by a core size dependent one-group criticality expression. Typical results are shown for a W-Re/UN cermet-fueled, lithium-cooled space reactor over a power range of 25 to 100 MWt. Reactor sensitivity coefficients are also shown for changes in reactor weight and number of coolant channels due to changes in core thermal/hydraulic constraints.

Specht, E.R.; Villalobos, A. (Rockwell International, Rocketdyne Division, 6633 Canoga Avenue, HB23, Canoga Park, California (USA))

1991-01-10T23:59:59.000Z

390

Refurbishment of the rotating rack of the OSU TRIGA MKII reactor  

SciTech Connect

Many TRIGA reactors have experienced operational difficulties with the rotating racks used for sample irradiation. Generally the rack gradually becomes more difficult to rotate until it finally seizes. The recommended action at that point is replacement of the entire facility at a significant cost. The purpose of this paper is to describe the symptoms leading to rack failure and to present the results of a refurbishment procedure that does not involve the use of solvents which create mixed chemical and radioactive hazardous waste. The primary reason for rack failure is the buildup of sludge produced through irradiation of lubrication oil. The refurbishment procedure involves using a commercially available degreasing solution which can be pumped into and out of the rack with the objective of removing this sludge. The solution used is sold under the trade name 'Simple Green'. No radioactive material was detected on smear or air samples taken of the work area during the reifurbishment activities and the rack rotates freely in both direction even after eighteen months of operation. The only disadvantage to performing this procedure has been the need to maintain a very aggressive contamination control program when unloading samples from the rack. A very fine particulate material attaches to the outside of tubes used to encapsulate samples. This material can produce contamination levels of 10,000 dpm/100 cm{sup 2} in the worst cases but will typically produce local hot spots on the order of 1000 dpm. (author)

Higginbotham, J.F.; Dodd, B.; Pratt, D.S.; Anderson, T.V. [OSU Radiation Center, Oregon State University, Corvallis, OR 97331-5903 (United States)

1992-07-01T23:59:59.000Z

391

Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230  

SciTech Connect

Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle Plates to the Baffle Former Plates. The FaST is designed to remove the Baffle Former Plates from the Core Barrel. The VRS further volume reduces segmented components using multiple configurations of the 38i and horizontal reciprocating saws. After the successful removal and volume reduction of the Internals, the RV will be segmented using a 'First in the US' thermal cutting process through a co-operative effort with Siempelkamp NIS Ingenieurgesellschaft mbH using their experience at the Stade NPP and Karlsruhe in Germany. SNS mobilized in the fall of 2011 to commence execution of the project in order to complete the RVI segmentation, removal and packaging activities for the first unit (Unit 2) by end of the 2012/beginning 2013 and then mobilize to the second unit, Unit 1. Parallel to the completion of the segmentation of the reactor vessel internals at Unit 1, SNS will segment the Unit 2 pressure vessel and at completion move to Unit 1. (authors)

Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)] [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

2013-07-01T23:59:59.000Z

392

Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who  

SciTech Connect

The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.

Forsberg, C.W.; Reich, W.J.

1991-09-01T23:59:59.000Z

393

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

SciTech Connect

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

Deitrich, L. W.

1998-07-28T23:59:59.000Z

394

Nuclear divisional reactor  

SciTech Connect

A nuclear divisional reactor including a reactor core having side and top walls, a heat exchanger substantially surrounding the core, the heat exchanger including a plurality of separate fluid holding and circulating chambers each in contact with a portion of the core, control rod means associated with the core and external of the heat exchanger including control rods and means for moving said control rods, each of the chambers having separate means for delivering and removing fluid therefrom, separate means associated with each of the delivering and removing means for producing useable energy external of the chambers, each of the means for producing useable energy having separate variable capacity energy outputs thereby making available a plurality of individual sources of useable energy of varying degrees.

Administratrix, A.P.; Rugh, J.L.

1982-11-02T23:59:59.000Z

395

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

396

A novel reactor configuration for packed bed chemical-looping combustion of syngas  

Science Journals Connector (OSTI)

Abstract This study reports on the application of chemical looping combustion (CLC) in pressurized packed bed reactors using syngas as a fuel. High pressure operation of CLC in packed bed has a different set of challenges in terms of material properties, cycle and reactor design compared to fluidized bed operation. However, high pressure operation allows the use of inherently more efficient power cycles than low pressure fluidized bed solutions. This paper quantifies the challenges in high pressure operation and introduces a novel reactor concept with which those challenges can be addressed. Continuous cyclic operation of a packed bed CLC system is simulated in a 1D numerical reactor model. Importantly, it is demonstrated that the temperature profiles that can occur in a packed bed reactor as a result of the different process steps do not accumulate, and have a negligible effect on the overall performance of the system. Moreover, it has been shown that an even higher energy efficiency can be achieved by feeding the syngas from the opposite direction during the reduction step (i.e. countercurrent operation). Unfortunately, in this configuration mode, more severe temperature fluctuations occur in the reactor exhaust, which is disadvantageous for the operation of a downstream gas turbine. Finally, a novel reactor configuration is introduced in which the desired temperature rise for obtained hot pressured air suitable for a gas turbine is obtained by carrying out the process with two packed bed reactor in series (two-stage CLC). This is shown to be a good alternative to the single bed configuration, and has the added advantage of decreasing the demands on both the oxygen carrier and the reactor materials and design specification.

H.P. Hamers; F. Gallucci; P.D. Cobden; E. Kimball; M. van Sint Annaland

2013-01-01T23:59:59.000Z

397

System modeling and reactor design studies of the Advanced Thermionic Initiative space nuclear reactor  

SciTech Connect

In-core thermionic space reactor design concepts that operate at a nominal power output range of 20 to 50 kW(electric) are described. Details of the neutronic, thermionic, thermal hydraulics, and shielding performance are presented. Because of the strong absorption of thermal neutrons by natural tungsten and the large amount of natural tungsten within the reactor core, two designs are considered. An overall system design code has been developed at Oregon State University to model advanced in-core thermionic energy conversion-based nuclear reactor systems for space applications. The results show that the driverless single-cell Advanced Thermionic Initiative (ATI) configuration, which does not have driver fuel rods, proved to be more efficient than the driven core, which has driver rods. The results also show that the inclusion of the true axial and radial power distribution decrease the overall conversion efficiency. The flattening of the radial power distribution by three different methods would lead to a higher efficiency. The results show that only one TFE works at the optimum emitter temperature; all other TFEs are off the optimum performance and result in a 40% decrease of the efficiency of the overall system. The true axial profile is significantly different as there is a considerable amount of neutron leakage out of the top and bottom of the reactor. The analysis reveals that the axial power profile actually has a chopped cosine shape. For this axial profile, the reactor core overall efficiency for the driverless ATI reactor version is found to be 5.84% with a total electrical power of 21.92 kW(electric). By considering the true axial power profile instead of the uniform power profile, each TFE loses {approximately}80 W(electric).

Lee, H.H.; Abdul-Hamid, S.; Klein, A.C. [Oregon State Univ., Corvallis, OR (United States). Dept. of Nuclear Engineering Radiation Center] [Oregon State Univ., Corvallis, OR (United States). Dept. of Nuclear Engineering Radiation Center

1996-07-01T23:59:59.000Z

398

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

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

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC)

399

High Flux Beam Reactor | Environmental Restoration Projects | BNL  

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

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

400

Metrology/viewing system for next generation fusion reactors  

SciTech Connect

Next generation fusion reactors require accurate measuring systems to verify sub-millimeter alignment of plasma-facing components in the reactor vessel. A metrology system capable of achieving such accuracy must be compatible with the vessel environment of high gamma radiation, high vacuum, elevated temperature, and magnetic field. This environment requires that the system must be remotely deployed. A coherent, frequency modulated laser radar system is being integrated with a remotely operated deployment system to meet these requirements. The metrology/viewing system consists of a compact laser transceiver optics module which is linked through fiber optics to the laser source and imaging units that are located outside of the harsh environment. The deployment mechanism is a telescopic-mast positioning system. This paper identifies the requirements for the International Thermonuclear Experimental Reactor metrology and viewing system, and describes a remotely operated precision ranging and surface mapping system.

Spampinato, P.T.; Barry, R.E.; Chesser, J.B.; Menon, M.M. [Oak Ridge National Lab., TN (United States); Dagher, M.A. [Boeing Rocketdyne Div., Canoga Park, CA (United States)

1997-02-01T23:59:59.000Z

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


401

Piezoelectric material for use in a nuclear reactor core  

SciTech Connect

In radiation environments ultrasonic nondestructive evaluation has great potential for improving reactor safety and furthering the understanding of radiation effects and materials. In both nuclear power plants and materials test reactors, elevated temperatures and high levels of radiation present challenges to ultrasonic NDE methodologies. The challenges are primarily due to the degradation of the ultrasonic sensors utilized. We present results from the operation of a ultrasonic piezoelectric transducer, composed of bulk single crystal AlN, in a nuclear reactor core for over 120 MWHrs. The transducer was coupled to an aluminum cylinder and operated in pulse echo mode throughout the irradiation. In addition to the pulse echo testing impedance data were obtained. Further, the piezoelectric coefficient d{sub 33} was measured prior to irradiation and found to be 5.5 pC/N which is unchanged from as-grown samples, and in fact higher than the measured d{sub 33} for many as-grown samples.

Parks, D. A.; Reinhardt, Brian; Tittmann, B. R. [EES Department, Penn State University, University Park, PA 16802 (United States)

2012-05-17T23:59:59.000Z

402

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

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

(RISMC) Advanced Test (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for

403

Design operators  

E-Print Network (OSTI)

Design operators is a thesis that investigates the nature and characteristics of the design process by examining the interaction of computation with architectural design. The effects of the introduction of these media in ...

Dritsas, Stylianos, 1978-

2004-01-01T23:59:59.000Z

404

Business Operations  

Office of Energy Efficiency and Renewable Energy (EERE)

The Office of Business Operations is the central organization for all Office of Energy Efficiency and Renewable Energy (EERE) business products, processes, and systems. The three main offices of...

405

Operating Costs  

Directives, Delegations, and Requirements

This chapter is focused on capital costs for conventional construction and environmental restoration and waste management projects and examines operating cost estimates to verify that all elements of the project have been considered and properly estimated.

1997-03-28T23:59:59.000Z

406

Worker exposure for at-reactor management of spent nuclear fuel  

Science Journals Connector (OSTI)

......Impact Statement for the proposed Yucca Mountain repository(6,7); potential...a No Action Alternative in the Yucca Mountain Environmental Impact Statement...operations at reactor sites. FUNDING This research was funded by the......

Philippe F. Weck

2013-09-01T23:59:59.000Z

407

Assessment of light water reactor power plant cost and ultra-acceleration depreciation financing  

E-Print Network (OSTI)

Although in many regions of the U.S. the least expensive electricity is generated from light-water reactor (LWR) plants, the fixed (capital plus operation and maintenance) cost has increased to the level where the cost ...

El-Magboub, Sadek Abdulhafid.

408

Workers Demolish Reactor Support Facility as Part of River Corridor Contract  

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

RICHLAND, Wash. – EM’s Richland Operations Office and cleanup contractor Washington Closure Hanford recently completed the cleanout and demolition of the last reactor support facility as part of the River Corridor Closure Contract.

409

Non-linear Dynamical Reliability Analysis in the Very High Temperature Gas Cooled Reactor  

Science Journals Connector (OSTI)

A dynamic safety assessment has been developed for the passive system in the very high temperature gas cooled reactor (VHTR), where the operational data are deficient. It is needed to make use of the character...

Taeho Woo

2012-01-01T23:59:59.000Z

410

Radiative energy exhaust by sputtered and seeded impurities in fusion reactor  

Science Journals Connector (OSTI)

The combine effect of the seeded and sputtered impurities on the power load to the divertor plate and operation of fusion reactor is investigated in the paper. Since the ... temperature). The steady states of ITE...

R. Stankiewicz

2006-10-01T23:59:59.000Z

411

Biological processing in oscillatory baffled reactors: operation, advantages and potential  

Science Journals Connector (OSTI)

...optimized with respect to the chemical environment [72]; however...saccharification) using various chemical or biological methods, thereby...fermented into a variety of useful chemicals including ethanol and lactic...for using OBRs as a generic platform for culture of facultative...

2013-01-01T23:59:59.000Z

412

On Operational Power Reactor Regime and Ignited Spherical Tokamaks  

E-Print Network (OSTI)

, 2003 version of the "cold" magnetic "Fusion without ignition" in the next 35 years, the talk.-Pitersburg, St.-Pitersburg, RF % Insutute of Nuclear Fusion, RRC "Kurchatov Ins.", Moscow, RF & Vyoptics, Inc for magnetic fusion, OPRR requires a low recycling and wall-stabilized high- plasma. Because of the small

Zakharov, Leonid E.

413

An assessment of the core degradation frequency in a typical large LMFBR design for internal accident initiators-a comparison with PWR predictions  

SciTech Connect

A comparative assessment of the core degradation frequency due to internal accident initiators between a typical large liquid-metal fast breeder reactor (LMFBR) design and pressurized water reactors (PWRs) has been performed. For the PWR system, existing analyses have been utilized. For the reference LMFBR, an extensive analysis has been performed for two accident initiators, i.e., loss of off-site power and loss of main feedwater. Based on this analysis an estimate of about1 X 10/sup -6//reactor X yr has been obtained for the core degradation frequency of the reference LMFBR. This estimate is significantly smaller than the PWR core degradation frequency ( about 6 X 10/sup -5//yr). A sensitivity analysis shows that the parameters having the largest impact on the unavailability of decay heat removal are (a) for the ''loss of off-site power'' initiator: human error and failure to restore off-site power, and (b) for the ''loss of main feedwater'' initiator: the leakage rates of the passive decay heat removal system and the adoption of the policy to repair the Na-NaK heat exchanger only during normal shutdowns. The results indicate that the LMFBR system has the potential of higher resistance than the PWR system to the accident initiators considered. The lower core degradation frequency estimated for the LMFBR system is due to the presence of two redundant and diverse reactor shutdown systems, with a self-actuated feature included in one of them, the incorporation of a passive decay heat removal system, and the significantly lower sensitivity of the reference LMFBR to primary system pipe breaks.

Tzanos, C.P.; Adamantiades, A.G.; Hanan, N.A.

1983-12-01T23:59:59.000Z

414

IGORR-IV -- Proceedings of the fourth meeting of the International Group on Research Reactors  

SciTech Connect

The International Group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Twenty-nine papers were presented in five sessions and written versions of the papers or hard copies of the vugraphs used are published in these proceedings. The five sessions were: (1) Operating Research Reactors and Facility Upgrades; (2) Research Reactors in Design and Construction; (3) ANS Closeout Activities; (4) and (5) Research, Development, and Analysis Results.

Rosenbalm, K.F. [comp.] [comp.

1995-12-31T23:59:59.000Z

415

Thermionic Reactor Design Studies  

SciTech Connect

During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full-core-height diodes. Moreover, placing the fuel on the outside of the diode makes possible reactors with much higher fuel volume fractions, which enable power-flattened fast reactors scalable to very low power levels without the need for life-limiting hydride moderators or the use of efficiency-limiting driver fuel. In addition, with the fuel on the outside its swelling does not increase the emitter diameter or reduce the interelectrode gap. This should permit long lifetimes even with closer spacings, which can significantly improve the system efficiences. This was confirmed by coupled neutronic, thermal, thermionic, and electrical system analyses - some of which are presented in this paper - and by subsequent experiments. A companion paper presented next describes the fabrication and testing of full-scale converter elements, both fueled and unfueled, and summarizes the test results obtained. There is a duplicate copy in the file.

Schock, Alfred

1994-06-01T23:59:59.000Z

416

Automatic coolant flow control device for a nuclear reactor assembly  

DOE Patents (OSTI)

A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

Hutter, Ernest (Wilmette, IL)

1986-01-01T23:59:59.000Z

417

Post-doc: Modelling & Control of Continuous Reactors  

E-Print Network (OSTI)

development in mechatronics and microsystems, sustainable industrial processes, transportation systems scope is a demonstration of continuous reactors with in-line analytics for fine chemical production for improvement of operation and actuation and will provide data for sensor development and control guidelines. 4

Langendoen, Koen

418

Automatic coolant flow control device for a nuclear reactor assembly  

DOE Patents (OSTI)

A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

Hutter, E.

1984-01-27T23:59:59.000Z

419

The results of systems tests of the 500 kV busbar controllable shunting reactor in the Tavricheskaya substation  

SciTech Connect

The results of systems tests of the 500 kV busbar magnetization-controllable shunting reactor (CSR), set up in the Tavricheskaya substation, including measurements of the quality of the electric power, the harmonic composition of the network currents of the reactor for different values of the reactive power consumed, the determination of the regulating characteristics of the reactor, the speed of response of the shunting reactor in the current and voltage stabilization modes, and also the operation of the reactor under dynamic conditions for different perturbations, are presented. The results obtained are analyzed.

Gusev, S. I. [JSC 'FSK EES' (Russian Federation); Karpov, V. N.; Kiselev, A. N.; Kochkin, V. I. [Scientific-Research Institute of Electric Power Engineering (VNIIE) - Branch of the JSC 'NTTs Elektroenergetiki', (Russian Federation)

2009-09-15T23:59:59.000Z

420

Steady-state inductive spheromak operation  

DOE Patents (OSTI)

The inductively formed spheromak configuration (S-1) can be maintained in a highly stable and controlled fashion. The method described eliminates the restriction to pulsed spheromak plasmas or the use of electrodes for steady-state operation, and, therefore, is a reactor-relevant formation and sustainment method.

Janos, A.C.; Jardin, S.C.; Yamada, M.

1985-02-20T23:59:59.000Z

Note: This page contains sample records for the topic "reactor operators typically" 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

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - I: DUPIC Fuel Fabrication Cost  

SciTech Connect

A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC ({approx}49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC.

Choi, Hangbok; Ko, Won Il; Yang, Myung Seung [Korea Atomic Energy Research Institute (Korea, Republic of)

2001-05-15T23:59:59.000Z

422

Sorption-desorption characteristics of uranium, cesium and strontium in typical podzol soils from Ukraine  

Science Journals Connector (OSTI)

......cesium and strontium in typical podzol soils from Ukraine S. Mishra 1 H. Arae 1 P. V. Zamostyan 2 T. Ishikawa...Radiation Medicine of Academy of Medical Sciences of Ukraine, Kiev, Ukraine Sorption-desorption behaviour of uranium (U......

S. Mishra; H. Arae; P. V. Zamostyan; T. Ishikawa; H. Yonehara; S. K. Sahoo

2012-11-01T23:59:59.000Z

423

Brain Bases of Reading Fluency in Typical Reading and Impaired Fluency in Dyslexia  

E-Print Network (OSTI)

Although the neural systems supporting single word reading are well studied, there are limited direct comparisons between typical and dyslexic readers of the neural correlates of reading fluency. Reading fluency deficits ...

Christodoulou, Joanna

424

E-Print Network 3.0 - andisol typic hapludand Sample Search Results  

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

considerable pressure difference (typically overcontent... :427-436. Sperry, J.S., V. Stiller, and U.G. Hacke. 2002b. Soil water uptake andisolated vascular bundles and whole...

425

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

Office of Environmental Management (EM)

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

426

Systems analysis of the CANDU 3 Reactor  

SciTech Connect

This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

1993-07-01T23:59:59.000Z

427

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.

428

Determination of a peak benzene exposure to consumers at typical self-service gasoline stations  

E-Print Network (OSTI)

DETERMINATION OF A PEAK BENZENE EXPOSURE TO CONSUMERS AT TYPICAL SELF-SERVICE GASOLINE STATIONS A Thesis by TED CARAPEZZA Submitted to the Graduate College of Texas A8M University in Partial fulfillment of the requirement for the degree... of MASTER OF SCIENCE December 1977 Major Subject: Industrial Hygiene DETERMINATION OF A PEAK BENZENE EXPOSURE TO CONSUMERS AT TYPICAL SELF-SERVICE GASOLINE STATIONS A Thesis by TED CARAPEZZA Approved as to style and content by: (. (iL, &? Chairman...

Carapezza, Ted

2012-06-07T23:59:59.000Z

429

The impact of sheared vs. sawn timber in the typical southern pine plywood mill  

E-Print Network (OSTI)

THE IMPACT OF SHEARED VS' SAWN TIMBER IN THE TYPICAL SOUTHERN PINE PLYWOOD MILL A Thesis by RUSSELL GARRETT SWINNEY Submitted to the Office of Graduate Studies of Texas ASM University in partial fulfillment of the requirements... for the degree of MASTER OF SCIENCE December 1989 Major Subject: Forestry THE IMPACT OP SHEARED VS. SAWN T1MBER IN THE TYPICAL SOUTHERN PINE PLYWOOD MILI A Thesis by RUSSELL GARRETT SWINNEY Approved as to style and content by: Jease ( hair of mmi Jy...

Swinney, Russell Garrett

1989-01-01T23:59:59.000Z

430

Development and transfer of fuel fabrication and utilization technology for research reactors  

SciTech Connect

Approximately 300 research reactors supplied with US-enriched uranium are currently in operation in about 40 countries, with a variety of types, sizes, experiment capabilities and applications. Despite the usefulness and popularity of research reactors, relatively few innovations in their core design have been made in the last fifteen years. The main reason can be better understood by reviewing briefly the history of research reactor fuel technology and enrichment levels. Stringent requirements on the enrichment of the uranium to be used in research reactors were considered and a program was launched to assist research reactors in continuing their operation with the new requirements and with minimum penalties. The goal of the new program, the Reduced Enrichment Research and Test Reactor (RERTR) Program, is to develop the technical means to utilize LEU instead of HEU in research reactors without significant penalties in experiment performance, operating costs, reactor modifications, and safety characteristics. This paper reviews briefly the RERTR Program activities with special emphasis on the technology transfer aspects of interest to this conference.

Travelli, A.; Domagala, R.F.; Matos, J.E.; Snelgrove, J.L.

1982-01-01T23:59:59.000Z

431

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary  

SciTech Connect

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

None

2013-09-25T23:59:59.000Z

432

BOREAS Operations  

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

Study Area Operations/Thompson Airport (NSA-Ops) Study Area Operations/Thompson Airport (NSA-Ops) NSA Operations (NSA-Ops) The Keewatin Air Hanger: site of BOREAS Ops 1994 Dr. Piers Sellers working in Ops, 1994 BOREAS "Air Force" The NASA C-130 The University of Wyoming King Air The NASA Helicopter The NRC Twin Otter The NCAR Electra The Ontario Chieftain Back to the BOREAS Photo Page Index Other Sites: NSA Photos ||NSA-BP Photos | NSA-Fen Photos | NSA-OA Photos | NSA-OBS Photos | NSA-OJP Photos | NSA-UBS Photos | NSA-YJP Photos | NSA-Ops Photos SSA Photos || SSA-Airport Photos | SSA-Fen Photos | SSA-Mix Photos | SSA-OA Photos | SSA-OBS Photos | SSA-OJP Photos | SSA-YA Photos | SSA-YJP Photos | SSA-Ops Photos | ORNL DAAC Home || ORNL Home || NASA || Privacy, Security, Notices || Data Citation || Rate Us || Help |

433

SSA Operations  

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

Area Operations (SSA-Ops) Area Operations (SSA-Ops) "BOREAS Ops" was located at the Snodrifters Lodge, in Candle Lake, Saskatchewan. Radiosonde balloon launch at Ops The NASA Helicopter lands at Ops A meeting at the Snodrifter's Lodge Release of a radiosonde at the SSA operations center in Candle Lake. Back to the BOREAS Photo Page Index Other Sites: NSA Photos ||NSA-BP Photos | NSA-Fen Photos | NSA-OA Photos | NSA-OBS Photos | NSA-OJP Photos | NSA-UBS Photos | NSA-YJP Photos | NSA-Ops Photos SSA Photos || SSA-Airport Photos | SSA-Fen Photos | SSA-Mix Photos | SSA-OA Photos | SSA-OBS Photos | SSA-OJP Photos | SSA-YA Photos | SSA-YJP Photos | SSA-Ops Photos | ORNL DAAC Home || ORNL Home || NASA || Privacy, Security, Notices || Data Citation || Rate Us || Help | User Services - Tel: +1 (865) 241-3952 or E-mail: uso@daac.ornl.gov

434

Zircaloy performance in light water reactors  

SciTech Connect

Zircaloy has been successfully used as the primary light water reactor (LWR) core structural material since its introduction in the early days of the US naval nuclear program. Its unique combination of low neutron absorption cross section, fabricability, mechanical strength, and corrosion resistance in water and steam near 300{degrees}C has resulted in remarkable reliability of operation of pressurized and boiling water reactor (PWR, BWR) fuel through the years. At present time, BWRs use Zircaloy-2 and PWRs use Zircaloy-4 for fuel cladding. In BWRs, both Zircaloy-2 and -4 have been successfully used for spacer grids and channels, and in PWRs Zircaloy-4 is used for spacer grids and control rod guide tubes. Performance of fuel rods has been excellent thus far. The current trend for utilities worldwide is to expect both higher fuel reliability in the future. Fuel suppliers have already achieved extended exposures in lead use assemblies, and have demonstrated excellent performance in all areas; therefore unsuspected problems are not likely to arise. However, as exposure and expectations continue to increase, Zircaloy is being taken toward the limits of its known capabilities. This paper reviews Zircaloy performance capabilities in areas related to environmentally affected microstructure, mechanical properties, corrosion resistance, and dimensional stability. The effects of radiation and reactor environment on each property is illustrated with data, micrographs, and analysis.

Adamson, R.B.; Cheng, B.C.; Kruger, R.M. [GE Nuclear Energy, Pleasanton, CA (United States)

1992-12-31T23:59:59.000Z

435

CASL milestone validates reactor model using TVA data | ornl.gov  

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

Ron Walli Ron Walli Communications 865.576.0226 CASL milestone validates reactor model using TVA data This CASL visualization shows the thermal distribution of neutrons in Watts Bar Unit 1 Cycle 1 reactor core at initial criticality, as calculated by the VERA program. Image courtesy of Oak Ridge National Laboratory. This CASL visualization shows the thermal distribution of neutrons in Watts Bar Unit 1 Cycle 1 reactor core at initial criticality, as calculated by the VERA program. Image courtesy of Oak Ridge National Laboratory. (hi-res image) OAK RIDGE, Tenn., July 10, 2013 -Today, the Consortium for Advanced Simulation of Light Water Reactors announced that its scientists have successfully completed their first full-scale simulation of an operating nuclear reactor. CASL is modeling nuclear reactors on supercomputers to

436

Effects of light water reactor coolant environment on the fatigue lives of  

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

Effects of light water reactor coolant environment on the fatigue lives of Effects of light water reactor coolant environment on the fatigue lives of reactor materials July 8, 2013 A metal component can become progressively degraded, and its structural integrity can be adversely impacted when it is subjected to repeated fluctuating loads, or fatigue loading. Fatigue loadings on nuclear reactor pressure vessel components can occur because of changes in pressure and temperature caused by transients during operation, such as reactor startup or shutdown and turbine trip events. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code recognizes fatigue as a possible cause of failure of reactor materials and provides rules for designing nuclear power plant components to avoid fatigue failures. For various materials, the ASME Code defines the

437

Fuel and cladding nano-technologies based solutions for long life heat-pipe based reactors  

SciTech Connect

A novel nuclear reactor concept, unifying the fuel pipe with fuel tube functionality has been developed. The structure is a quasi-spherical modular reactor, designed for a very long life. The reactor module unifies the fuel tube with the heat pipe and a graphite beryllium reflector. It also uses a micro-hetero-structure that allows the fission products to be removed in the heat pipe flow and deposited in a getter area in the cold zone of the heat pipe, but outside the neutron flux. The reactor operates as a breed and burn reactor - it contains the fuel pipe with a variable enrichment, starting from the hot-end of the pipe, meant to assure the initial criticality, and reactor start-up followed by area with depleted uranium or thorium that get enriched during the consumption of the first part of the enriched uranium. (authors)

Popa-Simil, L. [LAVM LLC, Los Alamos (United States)

2012-07-01T23:59:59.000Z

438

Overview of environmental control aspects for the gas-cooled fast reactor  

SciTech Connect

Environmental control aspects relating to release of radionuclides have been analyzed for the Gas-Cooled Fast Reactor (GCFR). Information on environmental control systems was obtained for the most recent GCFR designs, and was used to evaluate the adequacy of these systems. The GCFR has been designed by the General Atomic Company as an alternative to other fast breeder reactor designs, such as the Liquid Metal Fast Breeder Reactor (LMFBR). The GCFR design includes mixed oxide fuel and helium coolant. The environmental impact of expected radionuclide releases from normal operation of the GCFR was evaluated using estimated collective dose equivalent commitments resulting from 1 year of plant operation. The results were compared to equivalent estimates for the Light Water Reactor (LWR) and High-Temperature Gas-Cooled Reactor (HTGR). A discussion of uncertainties in system performances, tritium production rates, and radiation quality factors for tritium is included.

Nolan, A.M.

1981-05-01T23:59:59.000Z

439

Nuclear Reactor Materials and Fuels  

Science Journals Connector (OSTI)

Nuclear reactor materials and fuels can be classified into six categories: Nuclear fuel materials Nuclear clad materials Nuclear coolant materials Nuclear poison materials Nuclear moderator materials

Dr. James S. Tulenko

2012-01-01T23:59:59.000Z

440

Light Water Reactor Sustainability Newsletter  

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

hydraulics software RELAP-7 (which is under development in the Light Water Reactor Sustainability LWRS Program). A novel interaction between the probabilistic part (i.e., RAVEN)...

Note: This page contains sample records for the topic "reactor operators typically" 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.


441

Reactor Safety Research Programs Quarterly Report October - December 1981  

SciTech Connect

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

Edler, S. K.

1982-03-01T23:59:59.000Z

442

Advanced Reactors Thermal Energy Transport for Process Industries  

SciTech Connect

The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as liquid fuel production, district heating, desalination, hydrogen production, and other process heat applications, etc. Some of the major technology challenges that must be overcome before the advanced reactors could be licensed on the reactor side are qualification of next generation of nuclear fuel, materials that can withstand higher temperature, improvement in power cycle thermal efficiency by going to combined cycles, SCO2 cycles, successful demonstration of advanced compact heat exchangers in the prototypical conditions, and from the process side application the challenge is to transport the thermal energy from the reactor to the process plant with maximum efficiency (i.e., with minimum temperature drop). The main focus of this study is on doing a parametric study of efficient heat transport system, with different coolants (mainly, water, He, and molten salts) to determine maximum possible distance that can be achieved.

P. Sabharwall; S.J. Yoon; M.G. McKellar; C. Stoots; George Griffith

2014-07-01T23:59:59.000Z

443

Risk management at the Oak Ridge National Laboratory research reactors  

SciTech Connect

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

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

1994-12-31T23:59:59.000Z

444

Process simulation of refinery units including chemical reactors  

Science Journals Connector (OSTI)

Process simulation methods for design and operation of refinery units are well established as long as no chemical reactors are included. The feedstocks are divided into pseudo-components which enables calculation of phase equilibria and transport properties. When chemical reactors are present some chemical conversion takes place which obviously affects the nature of the pseudo-components and their properties. The stream leaving the reactor will not only be of a different composition than the stream entering the reactor but in addition, the pseudo-components making up the outlet stream will also have other physical properties than the ones in the inlet stream. These changes affect not only the reactor unit but also the simulation of the whole flow-sheet. The paper presents a detailed model for an adiabatic distillate hydrotreater which takes into account the elemental composition of the feed. A special simulation strategy has been developed to incorporate such reactor units into process simulators. Finally, the simulation strategy is illustrated for a hydrotreating plant.

Jens A. Hansen; Barry H. Cooper

1992-01-01T23:59:59.000Z

445

Compact Reversed-Field Pinch Reactors (CRFPR): preliminary engineering considerations  

SciTech Connect

The unique confinement physics of the Reversed-Field Pinch (RFP) projects to a compact, high-power-density fusion reactor that promises a significant reduction in the cost of electricity. The compact reactor also promises a factor-of-two reduction in the fraction of total cost devoted to the reactor plant equipment (i.e., fusion power core (FPC) plus support systems). In addition to operational and developmental benefits, these physically smaller systems can operate economically over a range of total power output. After giving an extended background and rationale for the compact fusion approaches, key FPC subsystems for the Compact RFP Reactor (CRFPR) are developed, designed, and integrated for a minimum-cost, 1000-MWe(net) system. Both the problems and promise of the compact, high-power-density fusion reactor are quantitatively evaluated on the basis of this conceptual design. The material presented in this report both forms a framework for a broader, more expanded conceptual design as well as suggests directions and emphases for related research and development.

Hagenson, R.L.; Krakowski, R.A.; Bathke, C.G.; Miller, R.L.; Embrechts, M.J.; Schnurr, N.M.; Battat, M.E.; LaBauve, R.J.; Davidson, J.W.

1984-08-01T23:59:59.000Z

446

Development of advanced strain diagnostic techniques for reactor environments.  

SciTech Connect

The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding. During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.

Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.; Hall, Aaron Christopher; Urrea, David Anthony,; Parma, Edward J.,

2013-02-01T23:59:59.000Z

447

Energy-Performance-Based Design-Build Process: Strategies for Procuring High-Performance Buildings on Typical Construction Budgets: Preprint  

SciTech Connect

NREL experienced a significant increase in employees and facilities on our 327-acre main campus in Golden, Colorado over the past five years. To support this growth, researchers developed and demonstrated a new building acquisition method that successfully integrates energy efficiency requirements into the design-build requests for proposals and contracts. We piloted this energy performance based design-build process with our first new construction project in 2008. We have since replicated and evolved the process for large office buildings, a smart grid research laboratory, a supercomputer, a parking structure, and a cafeteria. Each project incorporated aggressive efficiency strategies using contractual energy use requirements in the design-build contracts, all on typical construction budgets. We have found that when energy efficiency is a core project requirement as defined at the beginning of a project, innovative design-build teams can integrate the most cost effective and high performance efficiency strategies on typical construction budgets. When the design-build contract includes measurable energy requirements and is set up to incentivize design-build teams to focus on achieving high performance in actual operations, owners can now expect their facilities to perform. As NREL completed the new construction in 2013, we have documented our best practices in training materials and a how-to guide so that other owners and owner's representatives can replicate our successes and learn from our experiences in attaining market viable, world-class energy performance in the built environment.

Scheib, J.; Pless, S.; Torcellini, P.

2014-08-01T23:59:59.000Z

448

Analysis of kinetic models of the methanol-to-gasoline (MTG) process in an integral reactor  

Science Journals Connector (OSTI)

From experimental results obtained in a wide range of operating conditions (temperature and contact time) in an isothermal fixed bed integral reactor, the validity both of the kinetic models proposed in the literature as well as their modifications, for the methanol-to-gasoline (MTG) process at zero time on-stream, has been studied. The kinetic parameters for the various models have been calculated by solving the equation of mass conservation in the reactor for the lumps of the kinetic models. The usefulness of the model of Schipper and Krambeck for simulating the operation in the isothermal fixed bed integral reactor has been proven in the 573–648 K range.

Ana G. Gayubo; Pedro L. Benito; Andrés T. Aguayo; Itziar Aguirre; Javier Bilbao

1996-01-01T23:59:59.000Z

449

Reactor coolant pump flywheel  

DOE Patents (OSTI)

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

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

2013-11-26T23:59:59.000Z

450

Biparticle fluidized bed reactor  

DOE Patents (OSTI)

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

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

1995-04-25T23:59:59.000Z

451

Biparticle fluidized bed reactor  

DOE Patents (OSTI)

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

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

1996-02-27T23:59:59.000Z