DOE Office of Scientific and Technical Information (OSTI.GOV)

Krieg, R.

For future pressurized-water reactors, which should be designed against core-meltdown accidents, missiles generated inside the containment present a severe problem for its integrity. The masses and geometries of the missiles, as well as their velocities, may vary to a great extent. Therefore a reliable proof of the containment integrity is very difficult. In this article the potential sources of missiles are discussed, and the conclusion was reached that the generation of heavy missiles must be prevented. Steam explosions must not damage the reactor vessel head. Thus fragments of the head cannot become missiles that endanger the containment shell. Furthermore, duringmore » a melt-through failure of the reactor vessel under high pressure, the resulting forces must not catapult the whole vessel against the containment shell. Only missiles caused by hydrogen explosions may be tolerable, but shielding structures that protect the containment shell may be required. Further investigations are necessary. Finally, measures are described showing that the generation of heavy missiles can indeed be prevented. Investigations are currently being carried out that will confirm the strength of the reactor vessel head. In addition, a device for retaining the fragments of a failing reactor vessel is discussed.« less

Nuclear reactor containment structure with continuous ring tunnel at grade

DOEpatents

Seidensticker, Ralph W.; Knawa, Robert L.; Cerutti, Bernard C.; Snyder, Charles R.; Husen, William C.; Coyer, Robert G.

1977-01-01

A nuclear reactor containment structure which includes a reinforced concrete shell, a hemispherical top dome, a steel liner, and a reinforced-concrete base slab supporting the concrete shell is constructed with a substantial proportion thereof below grade in an excavation made in solid rock with the concrete poured in contact with the rock and also includes a continuous, hollow, reinforced-concrete ring tunnel surrounding the concrete shell with its top at grade level, with one wall integral with the reinforced concrete shell, and with at least the base of the ring tunnel poured in contact with the rock.

ENGINEERING AND CONSTRUCTING THE HALLAM NUCLEAR POWER FACILITY REACTOR STRUCTURE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mahlmeister, J E; Haberer, W V; Casey, D F

1960-12-15

The Hallam Nuclear Power Facility reactor structure, including the cavity liner, is described, and the design philosophy and special design requirements which were developed during the preliminary and final engineering phases of the project are explained. The structure was designed for 600 deg F inlet and 1000 deg F outlet operating sodium temperatures and fabricated of austenitic and ferritic stainless steels. Support for the reactor core components and adequate containment for biological safeguards were readily provided even though quite conservative design philosophy was used. The calculated operating characteristics, including heat generation, temperature distributions and stress levels for full-power operation, aremore » summarized. Ship fabrication and field installation experiences are also briefly related. Results of this project have established that the sodium graphite reactor permits practical and economical fabrication and field erection procedures; considerably higher operating design temperatures are believed possible without radical design changes. Also, larger reactor structures can be similarly constructed for higher capacity (300 to 1000 Mwe) nuclear power plants. (auth)« less

75 FR 28074 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-19

..., 2009, (74 FR 52829-52830). Wednesday, June 9, 2010, Conference Room T2-B1, Two White Flint North... Regulatory Guide (RG) 1.216, ``Containment Structural Integrity Evaluation for Internal Pressure Loadings... representatives of the NRC staff regarding draft final RG 1.216, ``Containment Structural Integrity Evaluation for...

Structure and creep of Russian reactor steels with a BCC structure

NASA Astrophysics Data System (ADS)

Sagaradze, V. V.; Kochetkova, T. N.; Kataeva, N. V.; Kozlov, K. A.; Zavalishin, V. A.; Vil'danova, N. F.; Ageev, V. S.; Leont'eva-Smirnova, M. V.; Nikitina, A. A.

2017-05-01

The structural phase transformations have been revealed and the characteristics of the creep and long-term strength at 650, 670, and 700°C and 60-140 MPa have been determined in six Russian reactor steels with a bcc structure after quenching and high-temperature tempering. Creep tests were carried out using specially designed longitudinal and transverse microsamples, which were fabricated from the shells of the fuel elements used in the BN-600 fast neutron reactor. It has been found that the creep rate of the reactor bcc steels is determined by the stability of the lath martensitic and ferritic structures in relation to the diffusion processes of recovery and recrystallization. The highest-temperature oxide-free steel contains the maximum amount of the refractory elements and carbides. The steel strengthened by the thermally stable Y-Ti nanooxides has a record high-temperature strength. The creep rate at 700°C and 100 MPa in the samples of this steel is lower by an order of magnitude and the time to fracture is 100 times greater than that in the oxide-free reactor steels.

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murata, K.K.; Williams, D.C.; Griffith, R.O.

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of themore » input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.« less

TERNARY ALLOY-CONTAINING PLUTONIUM

DOEpatents

Waber, J.T.

1960-02-23

Ternary alloys of uranium and plutonium containing as the third element either molybdenum or zirconium are reported. Such alloys are particularly useful as reactor fuels in fast breeder reactors. The alloy contains from 2 to 25 at.% of molybdenum or zirconium, the balance being a combination of uranium and plutonium in the ratio of from 1 to 9 atoms of uranlum for each atom of plutonium. These alloys are prepared by melting the constituent elements, treating them at an elevated temperature for homogenization, and cooling them to room temperature, the rate of cooling varying with the oomposition and the desired phase structure. The preferred embodiment contains 12 to 25 at.% of molybdenum and is treated by quenching to obtain a body centered cubic crystal structure. The most important advantage of these alloys over prior binary alloys of both plutonium and uranium is the lack of cracking during casting and their ready machinability.

Fluid-structure interaction in fast breeder reactors

NASA Astrophysics Data System (ADS)

Mitra, A. A.; Manik, D. N.; Chellapandi, P. A.

2004-05-01

A finite element model for the seismic analysis of a scaled down model of Fast breeder reactor (FBR) main vessel is proposed to be established. The reactor vessel, which is a large shell structure with a relatively thin wall, contains a large volume of sodium coolant. Therefore, the fluid structure interaction effects must be taken into account in the seismic design. As part of studying fluid-structure interaction, the fundamental frequency of vibration of a circular cylindrical shell partially filled with a liquid has been estimated using Rayleigh's method. The bulging and sloshing frequencies of the first four modes of the aforementioned system have been estimated using the Rayleigh-Ritz method. The finite element formulation of the axisymmetric fluid element with Fourier option (required due to seismic loading) is also presented.

MEANS FOR COOLING REACTORS

DOEpatents

Wheeler, J.A.

1957-11-01

A design of a reactor is presented in which the fuel elements may be immersed in a liquid coolant when desired without the necessity of removing them from the reactor structure. The fuel elements, containing the fissionable material are in plate form and are disposed within spaced slots in a moderator material, such as graphite to form the core. Adjacent the core is a tank containing the liquid coolant. The fuel elements are mounted in spaced relationship on a rotatable shaft which is located between the core and the tank so that by rotation of the shaft the fuel elements may be either inserted in the slots in the core to sustain a chain reaction or immersed in the coolant.

Catalytic reactor

DOEpatents

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

2009-03-10

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

A Wireless Monitoring System for Cracks on the Surface of Reactor Containment Buildings.

PubMed

Zhou, Jianguo; Xu, Yaming; Zhang, Tao

2016-06-14

Structural health monitoring with wireless sensor networks has been increasingly popular in recent years because of the convenience. In this paper, a real-time monitoring system for cracks on the surface of reactor containment buildings is presented. Customized wireless sensor networks platforms are designed and implemented with sensors especially for crack monitoring, which include crackmeters and temperature detectors. Software protocols like route discovery, time synchronization and data transfer are developed to satisfy the requirements of the monitoring system and stay simple at the same time. Simulation tests have been made to evaluate the performance of the system before full scale deployment. The real-life deployment of the crack monitoring system is carried out on the surface of reactor containment building in Daya Bay Nuclear Power Station during the in-service pressure test with 30 wireless sensor nodes.

Process and apparatus for adding and removing particles from pressurized reactors

DOEpatents

Milligan, John D.

1983-01-01

A method for adding and removing fine particles from a pressurized reactor is provided, which comprises connecting the reactor to a container, sealing the container from the reactor, filling the container with particles and a liquid material compatible with the reactants, pressurizing the container to substantially the reactor pressure, removing the seal between the reactor and the container, permitting particles to fall into or out of the reactor, and resealing the container from the reactor. An apparatus for adding and removing particles is also disclosed.

Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

DOEpatents

Gabor, John D.; Cassulo, John C.; Pedersen, Dean R.; Baker, Jr., Louis

1986-01-01

The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

DOEpatents

Gabor, John D.; Cassulo, John C.; Pedersen, Dean R.; Baker Jr., Louis

1986-07-01

The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

DOEpatents

Gabor, J.D.; Cassulo, J.C.; Pedersen, D.R.; Baker, L. Jr.

The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and can be discharged from the reactor core. The invention provides a porous bed of sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

Characterisation of well-adhered ZrO2 layers produced on structured reactors using the sonochemical sol-gel method

NASA Astrophysics Data System (ADS)

Jodłowski, Przemysław J.; Chlebda, Damian K.; Jędrzejczyk, Roman J.; Dziedzicka, Anna; Kuterasiński, Łukasz; Sitarz, Maciej

2018-01-01

The aim of this study was to obtain thin zirconium dioxide coatings on structured reactors using the sonochemical sol-gel method. The preparation method of metal oxide layers on metallic structures was based on the synergistic combination of three approaches: the application of ultrasonic irradiation during the synthesis of Zr sol-gel based on a precursor solution containing zirconium(IV) n-propoxide, the addition of stabilszing agents, and the deposition of ZrO2 on the metallic structures using the dip-coating method. As a result, dense, uniform zirconium dioxide films were obtained on the FeCrAlloy supports. The structured reactors were characterised by various physicochemical methods, such as BET, AFM, EDX, XRF, XRD, XPS and in situ Raman spectroscopy. The results of the structural analysis by Raman and XPS spectroscopy confirmed that the metallic surface was covered by a ZrO2 layer without any impurities. SEM/EDX mapping revealed that the deposited ZrO2 covered the metallic support uniformly. The mechanical and high temperature tests showed that the developed ultrasound assisted sol-gel method is an efficient way to obtain thin, well-adhered zirconium dioxide layers on the structured reactors. The prepared metallic supports covered with thin ZrO2 layers may be a good alternative to layered structured reactors in several dynamics flow processes, for example for gas exhaust abatement.

Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

DOEpatents

Hunsbedt, A.; Boardman, C.E.

1995-04-11

A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor is disclosed. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo`s structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated. 5 figures.

Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1995-01-01

A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo's structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated.

A Wireless Monitoring System for Cracks on the Surface of Reactor Containment Buildings

PubMed Central

Zhou, Jianguo; Xu, Yaming; Zhang, Tao

2016-01-01

Structural health monitoring with wireless sensor networks has been increasingly popular in recent years because of the convenience. In this paper, a real-time monitoring system for cracks on the surface of reactor containment buildings is presented. Customized wireless sensor networks platforms are designed and implemented with sensors especially for crack monitoring, which include crackmeters and temperature detectors. Software protocols like route discovery, time synchronization and data transfer are developed to satisfy the requirements of the monitoring system and stay simple at the same time. Simulation tests have been made to evaluate the performance of the system before full scale deployment. The real-life deployment of the crack monitoring system is carried out on the surface of reactor containment building in Daya Bay Nuclear Power Station during the in-service pressure test with 30 wireless sensor nodes. PMID:27314357

Steel Containment Vessel Model Test: Results and Evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Costello, J.F.; Hashimote, T.; Hessheimer, M.F.

A high pressure test of the steel containment vessel (SCV) model was conducted on December 11-12, 1996 at Sandia National Laboratories, Albuquerque, NM, USA. The test model is a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of an improved Mark II boiling water reactor (BWR) containment. A concentric steel contact structure (CS), installed over the SCV model and separated at a nominally uniform distance from it, provided a simplified representation of a reactor shield building in the actual plant. The SCV model and contact structure were instrumented with strain gages and displacement transducers to record the deformationmore » behavior of the SCV model during the high pressure test. This paper summarizes the conduct and the results of the high pressure test and discusses the posttest metallurgical evaluation results on specimens removed from the SCV model.« less

Nuclear reactor cavity floor passive heat removal system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluidmore » communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soldevilla, M.; Salmons, S.; Espinosa, B.

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a uniquemore » repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)« less

A preliminary survey of selected structures on the Hanford Site for Townsend`s big-eared bat (Plecotus townsendii)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Becker, J.M.

A preliminary survey of selected structures on the Hanford Site for Townsend`s big-wed bat (Plecotus townsendii) was conducted by Pacific Northwest Laboratory (PNL) in August and September 1993. The Westinghouse Hanford Company (WHC) commissioned PNL to evaluate the potential for this bat, a candidate for federal protection, to occur in buildings potentially affected by decontamination and decommissioning operations under the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). The project involved identifying structures that contained bats and determining whether Townsend`s big-eared bats were among those present. The survey focused on deactivated reactors, other buildings in the 100D and 100K Areas,more » canyon buildings in the 200 Areas, and other structures reported to contain bats. During this six-week survey, Townsend`s big-wed bat was not located. However, some structures likely to contain bat colonies were unable to be surveyed and others were only partially surveyed. These require further investigation over a longer period of time before a final determination on this species can be made. Of the buildings surveyed, the reactors and their associated buildings provided roosting sites most used by bats. No bats were found in canyon buildings in the 200 areas. These buildings are occupied, well-lighted, and offer few entrances for bats. They are also probably too distant from the Columbia River Shoreline, which constitutes the most important bat foraging habitat. We recommend that the remaining reactors and buildings, with emphasis on subterranean tunnels and basements, be surveyed during a more extended time period, i.e., June through September 1994.« less

Seismic attenuation system for a nuclear reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liszkai, Tamas; Cadell, Seth

A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vesselmore » are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.« less

Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beyer, Brian David; Beddingfield, David H; Durst, Philip

2010-01-01

The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguardsmore » criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.« less

Sealed head access area enclosure

DOEpatents

Golden, Martin P.; Govi, Aldo R.

1978-01-01

A liquid-metal-cooled fast breeder power reactor is provided with a sealed head access area enclosure disposed above the reactor vessel head consisting of a plurality of prefabricated structural panels including a center panel removably sealed into position with inflatable seals, and outer panels sealed into position with semipermanent sealant joints. The sealant joints are located in the joint between the edge of the panels and the reactor containment structure and include from bottom to top an inverted U-shaped strip, a lower layer of a room temperature vulcanizing material, a separator strip defining a test space therewithin, and an upper layer of a room temperature vulcanizing material. The test space is tapped by a normally plugged passage extending to the top of the enclosure for testing the seal or introducing a buffer gas thereinto.

High Efficiency Solar-based Catalytic Structure for CO 2 Reforming

DOE Office of Scientific and Technical Information (OSTI.GOV)

Menkara, Hisham

Throughout this project, we developed and optimized various photocatalyst structures for CO 2 reforming into hydrocarbon fuels and various commodity chemical products. We also built several closed-loop and continuous fixed-bed photocatalytic reactor system prototypes for a larger-scale demonstration of CO 2 reforming into hydrocarbons, mainly methane and formic acid. The results achieved have indicated that with each type of reactor and structure, high reforming yields can be obtained by refining the structural and operational conditions of the reactor, as well as by using various sacrificial agents (hole scavengers). We have also demonstrated, for the first time, that an aqueous solutionmore » containing acid whey (a common bio waste) is a highly effective hole scavenger for a solar-based photocatalytic reactor system and can help reform CO 2 into several products at once. The optimization tasks performed throughout the project have resulted in efficiency increase in our conventional reactors from an initial 0.02% to about 0.25%, which is 10X higher than our original project goal. When acid whey was used as a sacrificial agent, the achieved energy efficiency for formic acid alone was ~0.4%, which is 16X that of our original project goal and higher than anything ever reported for a solar-based photocatalytic reactor. Therefore, by carefully selecting sacrificial agents, it should be possible to reach energy efficiency in the range of the photosynthetic efficiency of typical crop and biofuel plants (1-3%).« less

Application of the TEMPEST computer code for simulating hydrogen distribution in model containment structures. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trent, D.S.; Eyler, L.L.

In this study several aspects of simulating hydrogen distribution in geometric configurations relevant to reactor containment structures were investigated using the TEMPEST computer code. Of particular interest was the performance of the TEMPEST turbulence model in a density-stratified environment. Computed results illustrated that the TEMPEST numerical procedures predicted the measured phenomena with good accuracy under a variety of conditions and that the turbulence model used is a viable approach in complex turbulent flow simulation.

Systems and methods for dismantling a nuclear reactor

DOEpatents

Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon

2014-10-28

Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

HOMOGENEOUS NUCLEAR REACTOR

DOEpatents

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

1959-02-17

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

75 FR 25303 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Regulatory...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-05-07

... (previously DG-1203) ``Containment Structural Integrity Evaluation for Internal Pressure Loadings Above Design... the time allotted to present oral statements can be obtained from the website cited above or by...

Spatial and Temporal Control of Chemical Structure for Biofouling Resistant, High Fouling Release Surfaces

DTIC Science & Technology

2014-06-02

thick-walled glass reactors fitted with ACE-threads under an argon atmosphere. The reactors were dried under vacuum then refilled with argon five times...were settled in individual dishes containing 10 ml of suspension at ~20°C on the laboratory benches. After 2 h the slides were exposed to a submerged ...and D20). This was done in order to see how the surfaces rearranged themselves after prolonged periods submerged in water, as would be the case on

Contact structure for use in catalytic distillation

DOEpatents

Jones, Jr., Edward M.

1984-01-01

A method for conducting catalytic chemical reactions and fractionation of the reaction mixture comprising feeding reactants into a distillation column reactor contracting said reactant in liquid phase with a fixed bed catalyst in the form of a contact catalyst structure consisting of closed porous containers containing the catatlyst for the reaction and a clip means to hold and support said containers, which are disposed above, i.e., on the distillation trays in the tower. The trays have weir means to provide a liquid level on the trays to substantially cover the containers. In other words, the trays function in their ordinary manner with the addition thereto of the catalyst. The reaction mixture is concurrently fractionated in the column.

Cylindrical micelles of a POSS amphiphilic dendrimer as nano-reactors for polymerization.

PubMed

Weng, Jing-Ting; Yeh, Tso-Fan; Samuel, Ashok Zachariah; Huang, Yi-Fan; Sie, Jyun-Hao; Wu, Kuan-Yi; Peng, Chi-How; Hamaguchi, Hiro-O; Wang, Chien-Lung

2018-02-15

A low generation amphiphilic dendrimer, POSS-AD, which has a POSS core and eight amphiphilic arms, was synthesized and used as a nano-reactor to produce well-defined polymer nano-cylinders. Confirmed by small-angle X-ray scattering (SAXS), Raman and NMR spectrometry, monodispersed cylindrical micelles that contain a hydrophilic cavity with a diameter of 2.09 nm and a length of 4.26 nm were produced via co-assembling POSS-AD with hydrophilic liquids, such as H 2 O and HEMA in hydrophobic solvents. Taking the HEMA/POSS-AD cylindrical micelles as nano-reactors, polymerization of HEMA within the micelles results in polymer nano-cylinders (POSS-ADNPs) with a diameter of 2.24 nm and a length of 5.02 nm. The study confirmed that despite the inability to maintain specific shape in solution, low generation dendrimers form well-defined nano-containers or nano-reactors, which relies on co-assembling with hydrophilic guest molecules. These nano-reactors are robust enough to maintain their shape during the polymerization of the guest molecules. Polymer nano-cylinders with dimensions less than 10 nm can thus be produced from the HEMA/POSS-AD micelles. Since the chemical structure of low-generation dendrimers and the contents of the co-assembled nano-reactors can be easily adjusted, the concept holds the potential for the further developments of low-generation amphiphilic dendrimers.

Removal design report for the 108-F Biological Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1997-09-01

Most of the 100-F facilities were deactivated with the reactor and have since been demolished. Of the dozen or so reactor-related structures, only the 105-F Reactor Building and the 108-F Biology Laboratory remain standing today. The 108-F Biology Laboratory was intended to be used as a facility for the mixing and addition of chemicals used in the treatment of the reactor cooling water. Shortly after F Reactor began operation, it was determined that the facility was not needed for this purpose. In 1949, the building was converted for use as a biological laboratory. In 1962, the lab was expanded bymore » adding a three-story annex to the original four-story structure. The resulting lab had a floor area of approximately 2,883 m{sup 2} (main building and annex) that operated until 1973. The building contained 47 laboratories, a number of small offices, a conference room, administrative section, lunch and locker rooms, and a heavily shielded, high-energy exposure cell. The purpose of this removal design report is to establish the methods of decontamination and decommissioning and the supporting functions associated with facility removal and disposal.« less

Investigation of materials for fusion power reactors

NASA Astrophysics Data System (ADS)

Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

2014-06-01

The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

Bubbling bed catalytic hydropyrolysis process utilizing larger catalyst particles and smaller biomass particles featuring an anti-slugging reactor

DOEpatents

Marker, Terry L; Felix, Larry G; Linck, Martin B; Roberts, Michael J

2014-09-23

This invention relates to a process for thermochemically transforming biomass or other oxygenated feedstocks into high quality liquid hydrocarbon fuels. In particular, a catalytic hydropyrolysis reactor, containing a deep bed of fluidized catalyst particles is utilized to accept particles of biomass or other oxygenated feedstocks that are significantly smaller than the particles of catalyst in the fluidized bed. The reactor features an insert or other structure disposed within the reactor vessel that inhibits slugging of the bed and thereby minimizes attrition of the catalyst. Within the bed, the biomass feedstock is converted into a vapor-phase product, containing hydrocarbon molecules and other process vapors, and an entrained solid char product, which is separated from the vapor stream after the vapor stream has been exhausted from the top of the reactor. When the product vapor stream is cooled to ambient temperatures, a significant proportion of the hydrocarbons in the product vapor stream can be recovered as a liquid stream of hydrophobic hydrocarbons, with properties consistent with those of gasoline, kerosene, and diesel fuel. Separate streams of gasoline, kerosene, and diesel fuel may also be obtained, either via selective condensation of each type of fuel, or via later distillation of the combined hydrocarbon liquid.

Bubbling bed catalytic hydropyrolysis process utilizinig larger catalyst particles and small biomass particles featuring an anti-slugging reactor

DOEpatents

Marker, Terry L.; Felix, Larry G.; Linck, Martin B.; Roberts, Michael J.

2016-12-06

This invention relates to a process for thermochemically transforming biomass or other oxygenated feedstocks into high quality liquid hydrocarbon fuels. In particular, a catalytic hydropyrolysis reactor, containing a deep bed of fluidized catalyst particles is utilized to accept particles of biomass or other oxygenated feedstocks that are significantly smaller than the particles of catalyst in the fluidized bed. The reactor features an insert or other structure disposed within the reactor vessel that inhibits slugging of the bed and thereby minimizes attrition of the catalyst. Within the bed, the biomass feedstock is converted into a vapor-phase product, containing hydrocarbon molecules and other process vapors, and an entrained solid char product, which is separated from the vapor stream after the vapor stream has been exhausted from the top of the reactor. When the product vapor stream is cooled to ambient temperatures, a significant proportion of the hydrocarbons in the product vapor stream can be recovered as a liquid stream of hydrophobic hydrocarbons, with properties consistent with those of gasoline, kerosene, and diesel fuel. Separate streams of gasoline, kerosene, and diesel fuel may also be obtained, either via selective condensation of each type of fuel, or via later distillation of the combined hydrocarbon liquid.

Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

NASA Technical Reports Server (NTRS)

Roth, R. J.

1976-01-01

The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peterson, Per F.

A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality ofmore » refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.« less

Flow dynamics in bioreactors containing tissue engineering scaffolds.

PubMed

Lawrence, Benjamin J; Devarapalli, Mamatha; Madihally, Sundararajan V

2009-02-15

Bioreactors are widely used in tissue engineering as a way to distribute nutrients within porous materials and provide physical stimulus required by many tissues. However, the fluid dynamics within the large porous structure are not well understood. In this study, we explored the effect of reactor geometry by using rectangular and circular reactors with three different inlet and outlet patterns. Geometries were simulated with and without the porous structure using the computational fluid dynamics software Comsol Multiphysics 3.4 and/or ANSYS CFX 11 respectively. Residence time distribution analysis using a step change of a tracer within the reactor revealed non-ideal fluid distribution characteristics within the reactors. The Brinkman equation was used to model the permeability characteristics with in the chitosan porous structure. Pore size was varied from 10 to 200 microm and the number of pores per unit area was varied from 15 to 1,500 pores/mm(2). Effect of cellular growth and tissue remodeling on flow distribution was also assessed by changing the pore size (85-10 microm) while keeping the number of pores per unit area constant. These results showed significant increase in pressure with reduction in pore size, which could limit the fluid flow and nutrient transport. However, measured pressure drop was marginally higher than the simulation results. Maximum shear stress was similar in both reactors and ranged approximately 0.2-0.3 dynes/cm(2). The simulations were validated experimentally using both a rectangular and circular bioreactor, constructed in-house. Porous structures for the experiments were formed using 0.5% chitosan solution freeze-dried at -80 degrees C, and the pressure drop across the reactor was monitored.

Heat dissipating nuclear reactor with metal liner

DOEpatents

Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

1985-11-21

A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

Heat dissipating nuclear reactor with metal liner

DOEpatents

Gluekler, Emil L.; Hunsbedt, Anstein; Lazarus, Jonathan D.

1987-01-01

Disclosed is a nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

Radiant vessel auxiliary cooling system

DOEpatents

Germer, John H.

1987-01-01

In a modular liquid-metal pool breeder reactor, a radiant vessel auxiliary cooling system is disclosed for removing the residual heat resulting from the shutdown of a reactor by a completely passive heat transfer system. A shell surrounds the reactor and containment vessel, separated from the containment vessel by an air passage. Natural circulation of air is provided by air vents at the lower and upper ends of the shell. Longitudinal, radial and inwardly extending fins extend from the shell into the air passage. The fins are heated by radiation from the containment vessel and convect the heat to the circulating air. Residual heat from the primary reactor vessel is transmitted from the reactor vessel through an inert gas plenum to a guard or containment vessel designed to contain any leaking coolant. The containment vessel is conventional and is surrounded by the shell.

Apparatus and process for the surface treatment of carbon fibers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paulauskas, Felix Leonard; Ozcan, Soydan; Naskar, Amit K.

A method for surface treating a carbon-containing material in which carbon-containing material is reacted with decomposing ozone in a reactor (e.g., a hollow tube reactor), wherein a concentration of ozone is maintained throughout the reactor by appropriate selection of at least processing temperature, gas stream flow rate, reactor dimensions, ozone concentration entering the reactor, and position of one or more ozone inlets (ports) in the reactor, wherein the method produces a surface-oxidized carbon or carbon-containing material, preferably having a surface atomic oxygen content of at least 15%. The resulting surface-oxidized carbon material and solid composites made therefrom are also described.

Contact structure for use in catalytic distillation

DOEpatents

Jones, E.M. Jr.

1984-03-27

A method is described for conducting catalytic chemical reactions and fractionation of the reaction mixture comprising feeding reactants into a distillation column reactor, contracting said reactant in liquid phase with a fixed bed catalyst in the form of a contact catalyst structure consisting of closed porous containers containing the catalyst for the reaction and a clip means to hold and support said containers, which are disposed above, i.e., on the distillation trays in the tower. The trays have weir means to provide a liquid level on the trays to substantially cover the containers. In other words, the trays function in their ordinary manner with the addition thereto of the catalyst. The reaction mixture is concurrently fractionated in the column. 7 figs.

Contact structure for use in catalytic distillation

DOEpatents

Jones, Jr., Edward M.

1985-01-01

A method and apparatus for conducting catalytic chemical reactions and fractionation of the reaction mixture, comprising and feeding reactants into a distillation column reactor contracting said reactant in a liquid phase with a fixed bed catalyst in the form of a contact catalyst structure, consisting of closed porous containers containing the catalyst for the reaction and a clip means to hold and support said containers, which are disposed above, i.e., on the distillation trays in the tower. The trays have weir means to provide a liquid level on the trays to substantially cover the containers. In other words, the trays function in their ordinary manner with the addition thereto of the catalyst. The reaction mixture is concurrently fractionated in the column.

Contact structure for use in catalytic distillation

DOEpatents

Jones, E.M. Jr.

1985-08-20

A method and apparatus are disclosed for conducting catalytic chemical reactions and fractionation of the reaction mixture, comprising and feeding reactants into a distillation column reactor contracting said reactant in a liquid phase with a fixed bed catalyst in the form of a contact catalyst structure, consisting of closed porous containers containing the catalyst for the reaction and a clip means to hold and support said containers, which are disposed above, i.e., on the distillation trays in the tower. The trays have weir means to provide a liquid level on the trays to substantially cover the containers. In other words, the trays function in their ordinary manner with the addition thereto of the catalyst. The reaction mixture is concurrently fractionated in the column. 7 figs.

SAFSIM theory manual: A computer program for the engineering simulation of flow systems

NASA Astrophysics Data System (ADS)

Dobranich, Dean

1993-12-01

SAFSIM (System Analysis Flow SIMulator) is a FORTRAN computer program for simulating the integrated performance of complex flow systems. SAFSIM provides sufficient versatility to allow the engineering simulation of almost any system, from a backyard sprinkler system to a clustered nuclear reactor propulsion system. In addition to versatility, speed and robustness are primary SAFSIM development goals. SAFSIM contains three basic physics modules: (1) a fluid mechanics module with flow network capability; (2) a structure heat transfer module with multiple convection and radiation exchange surface capability; and (3) a point reactor dynamics module with reactivity feedback and decay heat capability. Any or all of the physics modules can be implemented, as the problem dictates. SAFSIM can be used for compressible and incompressible, single-phase, multicomponent flow systems. Both the fluid mechanics and structure heat transfer modules employ a one-dimensional finite element modeling approach. This document contains a description of the theory incorporated in SAFSIM, including the governing equations, the numerical methods, and the overall system solution strategies.

An easily regenerable enzyme reactor prepared from polymerized high internal phase emulsions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ruan, Guihua, E-mail: guihuaruan@hotmail.com; Guangxi Collaborative Innovation Center for Water Pollution Control and Water Safety in Karst Area, Guilin University of Technology, Guilin 541004; Wu, Zhenwei

A large-scale high-efficient enzyme reactor based on polymerized high internal phase emulsion monolith (polyHIPE) was prepared. First, a porous cross-linked polyHIPE monolith was prepared by in-situ thermal polymerization of a high internal phase emulsion containing styrene, divinylbenzene and polyglutaraldehyde. The enzyme of TPCK-Trypsin was then immobilized on the monolithic polyHIPE. The performance of the resultant enzyme reactor was assessed according to the conversion ability of N{sub α}-benzoyl-L-arginine ethyl ester to N{sub α}-benzoyl-L-arginine, and the protein digestibility of bovine serum albumin (BSA) and cytochrome (Cyt-C). The results showed that the prepared enzyme reactor exhibited high enzyme immobilization efficiency and fast andmore » easy-control protein digestibility. BSA and Cyt-C could be digested in 10 min with sequence coverage of 59% and 78%, respectively. The peptides and residual protein could be easily rinsed out from reactor and the reactor could be regenerated easily with 4 M HCl without any structure destruction. Properties of multiple interconnected chambers with good permeability, fast digestion facility and easily reproducibility indicated that the polyHIPE enzyme reactor was a good selector potentially applied in proteomics and catalysis areas. - Graphical abstract: Schematic illustration of preparation of hypercrosslinking polyHIPE immobilized enzyme reactor for on-column protein digestion. - Highlights: • A reactor was prepared and used for enzyme immobilization and continuous on-column protein digestion. • The new polyHIPE IMER was quite suit for protein digestion with good properties. • On-column digestion revealed that the IMER was easy regenerated by HCl without any structure destruction.« less

Reactor core isolation cooling system

DOEpatents

Cooke, F.E.

1992-12-08

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

Reactor core isolation cooling system

DOEpatents

Cooke, Franklin E.

1992-01-01

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

Staged depressurization system

DOEpatents

Schulz, T.L.

1993-11-02

A nuclear reactor having a reactor vessel disposed in a containment shell is depressurized in stages using depressurizer valves coupled in fluid communication with the coolant circuit. At least one sparger submerged in the in-containment refueling water storage tank which can be drained into the containment sump communicates between one or more of the valves and an inside of the containment shell. The depressurizer valves are opened in stages, preferably at progressively lower coolant levels and for opening progressively larger flowpaths to effect depressurization through a number of the valves in parallel. The valves can be associated with a pressurizer tank in the containment shell, coupled to a coolant outlet of the reactor. At least one depressurization valve stage openable at a lowest pressure is coupled directly between the coolant circuit and the containment shell. The reactor is disposed in the open sump in the containment shell, and a further valve couples the open sump to a conduit coupling the refueling water storage tank to the coolant circuit for adding water to the coolant circuit, whereby water in the containment shell can be added to the reactor from the open sump. 4 figures.

Staged depressurization system

DOEpatents

Schulz, Terry L.

1993-01-01

A nuclear reactor having a reactor vessel disposed in a containment shell is depressurized in stages using depressurizer valves coupled in fluid communication with the coolant circuit. At least one sparger submerged in the in-containment refueling water storage tank which can be drained into the containment sump communicates between one or more of the valves and an inside of the containment shell. The depressurizer valves are opened in stages, preferably at progressively lower coolant levels and for opening progressively larger flowpaths to effect depressurization through a number of the valves in parallel. The valves can be associated with a pressurizer tank in the containment shell, coupled to a coolant outlet of the reactor. At least one depressurization valve stage openable at a lowest pressure is coupled directly between the coolant circuit and the containment shell. The reactor is disposed in the open sump in the containment shell, and a further valve couples the open sump to a conduit coupling the refueling water storage tank to the coolant circuit for adding water to the coolant circuit, whereby water in the containment shell can be added to the reactor from the open sump.

Passive filtration of air egressing from nuclear containment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Malloy, III, John D

2017-09-26

A nuclear reactor includes a reactor core comprising fissile material disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor. A containment compartment contains the radiological containment. A heat sink includes a chimney configured to develop an upward-flowing draft in response to heated fluid flowing into a lower portion of the chimney. A fluid conduit is arranged to receive fluid from the containment compartment and to discharge into the chimney. A filter may be provided, with the fluid conduit including a first fluid conduit arranged to receive fluid from the containment compartment and to discharge into anmore » inlet of the filter, and a second fluid conduit arranged to receive fluid from an outlet of the filter and to discharge into the chimney. As the draft is developed passively, there is no need for a blower or pump configured to move fluid through the fluid conduit.« less

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2013 CFR

2013-01-01

... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the...

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2014 CFR

2014-01-01

... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the...

An easily regenerable enzyme reactor prepared from polymerized high internal phase emulsions.

PubMed

Ruan, Guihua; Wu, Zhenwei; Huang, Yipeng; Wei, Meiping; Su, Rihui; Du, Fuyou

2016-04-22

A large-scale high-efficient enzyme reactor based on polymerized high internal phase emulsion monolith (polyHIPE) was prepared. First, a porous cross-linked polyHIPE monolith was prepared by in-situ thermal polymerization of a high internal phase emulsion containing styrene, divinylbenzene and polyglutaraldehyde. The enzyme of TPCK-Trypsin was then immobilized on the monolithic polyHIPE. The performance of the resultant enzyme reactor was assessed according to the conversion ability of Nα-benzoyl-l-arginine ethyl ester to Nα-benzoyl-l-arginine, and the protein digestibility of bovine serum albumin (BSA) and cytochrome (Cyt-C). The results showed that the prepared enzyme reactor exhibited high enzyme immobilization efficiency and fast and easy-control protein digestibility. BSA and Cyt-C could be digested in 10 min with sequence coverage of 59% and 78%, respectively. The peptides and residual protein could be easily rinsed out from reactor and the reactor could be regenerated easily with 4 M HCl without any structure destruction. Properties of multiple interconnected chambers with good permeability, fast digestion facility and easily reproducibility indicated that the polyHIPE enzyme reactor was a good selector potentially applied in proteomics and catalysis areas. Copyright © 2016 Elsevier Inc. All rights reserved.

High aspect ratio catalytic reactor and catalyst inserts therefor

DOEpatents

Lin, Jiefeng; Kelly, Sean M.

2018-04-10

The present invention relates to high efficient tubular catalytic steam reforming reactor configured from about 0.2 inch to about 2 inch inside diameter high temperature metal alloy tube or pipe and loaded with a plurality of rolled catalyst inserts comprising metallic monoliths. The catalyst insert substrate is formed from a single metal foil without a central supporting structure in the form of a spiral monolith. The single metal foil is treated to have 3-dimensional surface features that provide mechanical support and establish open gas channels between each of the rolled layers. This unique geometry accelerates gas mixing and heat transfer and provides a high catalytic active surface area. The small diameter, high aspect ratio tubular catalytic steam reforming reactors loaded with rolled catalyst inserts can be arranged in a multi-pass non-vertical parallel configuration thermally coupled with a heat source to carry out steam reforming of hydrocarbon-containing feeds. The rolled catalyst inserts are self-supported on the reactor wall and enable efficient heat transfer from the reactor wall to the reactor interior, and lower pressure drop than known particulate catalysts. The heat source can be oxygen transport membrane reactors.

Shutdown system for a nuclear reactor

DOEpatents

Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

1984-06-05

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

Shutdown system for a nuclear reactor

DOEpatents

Groh, Edward F.; Olson, Arne P.; Wade, David C.; Robinson, Bryan W.

1984-01-01

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion.

Control system for a small fission reactor

DOEpatents

Burelbach, James P.; Kann, William J.; Saiveau, James G.

1986-01-01

A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired. In another embodiment, a plurality of flexible hollow tubes each containing a neutron absorber are positioned adjacent to one another in spaced relation around the periphery of the reactor vessel and inside the outer neutron reflector with reactivity controlled by the extension and compression of all or some of the coiled hollow tubes. Yet another embodiment of the invention envisions the neutron reflector in the form of an expandable coil spring positioned in an annular space between the reactor vessel and an outer neutron absorbing structure for controlling the neutron flux reflected back into the reactor vessel.

Extending the maximum operation time of the MNSR reactor.

PubMed

Dawahra, S; Khattab, K; Saba, G

2016-09-01

An effective modification to extend the maximum operation time of the Miniature Neutron Source Reactor (MNSR) to enhance the utilization of the reactor has been tested using the MCNP4C code. This modification consisted of inserting manually in each of the reactor inner irradiation tube a chain of three polyethylene-connected containers filled of water. The total height of the chain was 11.5cm. The replacement of the actual cadmium absorber with B(10) absorber was needed as well. The rest of the core structure materials and dimensions remained unchanged. A 3-D neutronic model with the new modifications was developed to compare the neutronic parameters of the old and modified cores. The results of the old and modified core excess reactivities (ρex) were: 3.954, 6.241 mk respectively. The maximum reactor operation times were: 428, 1025min and the safety reactivity factors were: 1.654 and 1.595 respectively. Therefore, a 139% increase in the maximum reactor operation time was noticed for the modified core. This increase enhanced the utilization of the MNSR reactor to conduct a long time irradiation of the unknown samples using the NAA technique and increase the amount of radioisotope production in the reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

Molten metal reactor and method of forming hydrogen, carbon monoxide and carbon dioxide using the molten alkaline metal reactor

DOEpatents

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

2012-11-13

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.

Impact of single walled carbon nanotubes (SWNTs) on wastewater microbial communities

NASA Astrophysics Data System (ADS)

Goyal, Deepankar

Aim: Carbon nanotubes (CNTs) hold great promise in advancing our future, with potential applications such as adsorbents, conductive composites, energy storage devices, and more. Despite of numerous potential applications of CNTs, almost nothing so far is known about how such carbon-based nanomaterials would in future impact environmental processes such as wastewater treatment. The objective of the current study was to evaluate the impact of single-walled carbon nanotubes (SWNTs) on microbial communities and wastewater treatment processes in activated sludge bioreactors. Method: Closed system batch-scale reactors were used to simulate the activated sludge process. Two sets of triplicate reactors were analyzed to determine the effects of SWNTs and associated impurities compared to control reactors that contained no CNTs. Sub-samples for microbial community analyses were aseptically removed periodically from the bioreactors every ˜1 hour 15 minutes and held at -80°C until analyzed. Genomic DNA was extracted from bioreactor samples, and molecular profiles of the bacterial communities were determined using automated ribosomal intergenic spacer analysis (ARISA). The clones for the ARISA profiles having distinct ARISA peaks were picked and sequenced. Result: ARISA profiles revealed adverse changes in CNT-exposed bacterial communities compared to control reactors associated with CNTs. The phylogenetic analysis of cloned insert containing Internal Transcribed Spacer (ITS) region plus the 16S rRNA genes identified them belonging to taxonomic groups of the families Sphingomonadaceae and Cytophagacaceae , and the genus Zoogloea. Changes in community structure were observed in both SWNT-exposed and control reactors over the experimental time period. Also the date on which activated sludge was obtained from a wastewater treatment plant facility seemed to play a critical role in changing the community structure altogether, indicating the importance of analyzing microbial community structure in addition to abiotic and bulk biological factors when evaluating the potential impact of contaminants. In parallel to this study, another peer group made simultaneous measurements of chemical oxygen demand (COD) and abiotic factors, which indicated that while SWNTs adsorbed organic carbon (measured as COD), they did not significantly impact its degradation by the microbial community. Conclusion: These results indicate that SWNTs differentially impact members of the activated sludge reactor bacterial community. Significance of the Study: The finding that community structure was affected by SWNTs indicates that this emerging contaminant differentially impacted members of the activated sludge bacterial community, even on a short time period of exposure. These results raise the concern of SWNT impact on biological functions carried out by the activated sludge process and a question about their environmental impact in coming era of nanotechnology.

Integrated head package cable carrier for a nuclear power plant

DOEpatents

Meuschke, Robert E.; Trombola, Daniel M.

1995-01-01

A cabling arrangement is provided for a nuclear reactor located within a containment. Structure inside the containment is characterized by a wall having a near side surrounding the reactor vessel defining a cavity, an operating deck outside the cavity, a sub-space below the deck and on a far side of the wall spaced from the near side, and an operating area above the deck. The arrangement includes a movable frame supporting a plurality of cables extending through the frame, each connectable at a first end to a head package on the reactor vessel and each having a second end located in the sub-space. The frame is movable, with the cables, between a first position during normal operation of the reactor when the cables are connected to the head package, located outside the sub-space proximate the head package, and a second position during refueling when the cables are disconnected from the head package, located in the sub-space. In a preferred embodiment, the frame straddles the top of the wall in a substantially horizontal orientation in the first position, pivots about an end distal from the head package to a substantially vertically oriented intermediate position, and is guided, while remaining about vertically oriented, along a track in the sub-space to the second position.

Structural mechanics simulations

NASA Technical Reports Server (NTRS)

Biffle, Johnny H.

1992-01-01

Sandia National Laboratory has a very broad structural capability. Work has been performed in support of reentry vehicles, nuclear reactor safety, weapons systems and components, nuclear waste transport, strategic petroleum reserve, nuclear waste storage, wind and solar energy, drilling technology, and submarine programs. The analysis environment contains both commercial and internally developed software. Included are mesh generation capabilities, structural simulation codes, and visual codes for examining simulation results. To effectively simulate a wide variety of physical phenomena, a large number of constitutive models have been developed.

Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

NASA Astrophysics Data System (ADS)

Tyapkov, V. F.

2014-07-01

The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

Methods and apparatuses for the development of microstructured nuclear fuels

DOEpatents

Jarvinen, Gordon D [Los Alamos, NM; Carroll, David W [Los Alamos, NM; Devlin, David J [Santa Fe, NM

2009-04-21

Microstructured nuclear fuel adapted for nuclear power system use includes fissile material structures of micrometer-scale dimension dispersed in a matrix material. In one method of production, fissile material particles are processed in a chemical vapor deposition (CVD) fluidized-bed reactor including a gas inlet for providing controlled gas flow into a particle coating chamber, a lower bed hot zone region to contain powder, and an upper bed region to enable powder expansion. At least one pneumatic or electric vibrator is operationally coupled to the particle coating chamber for causing vibration of the particle coater to promote uniform powder coating within the particle coater during fuel processing. An exhaust associated with the particle coating chamber and can provide a port for placement and removal of particles and powder. During use of the fuel in a nuclear power reactor, fission products escape from the fissile material structures and come to rest in the matrix material. After a period of use in a nuclear power reactor and subsequent cooling, separation of the fissile material from the matrix containing the embedded fission products will provide an efficient partitioning of the bulk of the fissile material from the fission products. The fissile material can be reused by incorporating it into new microstructured fuel. The fission products and matrix material can be incorporated into a waste form for disposal or processed to separate valuable components from the fission products mixture.

Integrity assessment of grouted posttensioning cables and reinforced concrete of a nuclear containment building

NASA Astrophysics Data System (ADS)

Philipose, K.; Shenton, B.

2011-04-01

The Containment Buildings of CANDU Nuclear Generating Stations were designed to house nuclear reactors and process equipment and also to provide confinement of releases from a potential nuclear accident such as a Loss Of Coolant Accident (LOCA). To meet this design requirement, a post-tensioning system was designed to induce compressive stresses in the structure to counteract the internal design pressure. The CANDU reactor building at Gentilly-1 (G-1), Quebec, Canada (250 MWe) was built in the early 1970s and is currently in a decommissioned state. The structure at present is under surveillance and monitoring. In the year 2000, a field investigation was conducted as part of a condition assessment and corrosion was detected in some of the grouted post-tension cable strands. However, no further work was done at that time to determine the cause, nature, impact and extent of the corrosion. An investigation of the Gentilly-1 containment building is currently underway to assess the condition of grouted post-tensioning cables and reinforced concrete. At two selected locations, concrete and steel reinforcements were removed from the containment building wall to expose horizontal cables. Individual cable strands and reinforcement bars were instrumented and measurements were taken in-situ before removing them for forensic examination and destructive testing to determine the impact of ageing and corrosion. Concrete samples were also removed and tested in a laboratory. The purpose of the field investigation and laboratory testing, using this structure as a test bed, was also to collect material ageing data and to develop potential Nondestructive Examination (NDE) methods to monitor Containment Building Integrity. The paper describes the field work conducted and the test results obtained for concrete, reinforcement and post-tensioning cables.

Decommissioning of the High Flux Beam Reactor at Brookhaven National Laboratory.

PubMed

Hu, Jih-Perng; Reciniello, Richard N; Holden, Norman E

2012-08-01

The High Flux Beam Reactor (HFBR) at the Brookhaven National Laboratory was a heavy-water cooled and moderated reactor that achieved criticality on 31 October 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost 3 y for safety and environmental reviews. In November 1999, the United States Department of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR's cleanup performed during 1999-2009, to allow the BNL facilities to be re-accessed by the public, will be described in the paper.

Novel durable bio-photocatalyst purifiers, a non-heterogeneous mechanism: accelerated entrapped dye degradation into structural polysiloxane-shield nano-reactors.

PubMed

Dastjerdi, Roya; Montazer, Majid; Shahsavan, Shadi; Böttcher, Horst; Moghadam, M B; Sarsour, Jamal

2013-01-01

This research has designed innovative Ag/TiO(2) polysiloxane-shield nano-reactors on the PET fabric to develop novel durable bio-photocatalyst purifiers. To create these very fine nano-reactors, oppositely surface charged multiple size nanoparticles have been applied accompanied with a crosslinkable amino-functionalized polysiloxane (XPs) emulsion. Investigation of photocatalytic dye decolorization efficiency revealed a non-heterogeneous mechanism including an accelerated degradation of entrapped dye molecules into the structural polysiloxane-shield nano-reactors. In fact, dye molecules can be adsorbed by both Ag and XPs due to their electrostatic interactions and/or even via forming a complex with them especially with silver NPs. The absorbed dye and active oxygen species generated by TiO(2) were entrapped by polysiloxane shelter and the presence of silver nanoparticles further attract the negative oxygen species closer to the adsorbed dye molecules. In this way, the dye molecules are in close contact with concentrated active oxygen species into the created nano-reactors. This provides an accelerated degradation of dye molecules. This non-heterogeneous mechanism has been detected on the sample containing all of the three components. Increasing the concentration of Ag and XPs accelerated the second step beginning with an enhanced rate. Further, the treated samples also showed an excellent antibacterial activity. Copyright © 2012 Elsevier B.V. All rights reserved.

Neutronic calculation of fast reactors by the EUCLID/V1 integrated code

NASA Astrophysics Data System (ADS)

Koltashev, D. A.; Stakhanova, A. A.

2017-01-01

This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.

Chip-based device for parallel sorting, amplification, detection, and identification of nucleic acid subsequences

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beer, Neil Reginald; Colston, Jr, Billy W.

An apparatus for chip-based sorting, amplification, detection, and identification of a sample having a planar substrate. The planar substrate is divided into cells. The cells are arranged on the planar substrate in rows and columns. Electrodes are located in the cells. A micro-reactor maker produces micro-reactors containing the sample. The micro-reactor maker is positioned to deliver the micro-reactors to the planar substrate. A microprocessor is connected to the electrodes for manipulating the micro-reactors on the planar substrate. A detector is positioned to interrogate the sample contained in the micro-reactors.

Effect of reactor radiation on the thermal conductivity of TREAT fuel

NASA Astrophysics Data System (ADS)

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Wright, Arthur E.; Yacout, Abdellatif M.

2017-04-01

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO2 particle size on fission-fragment damage. The proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

Treatment of wastewater containing a large amount of suspended solids by a novel multi-staged UASB reactor.

PubMed

Uemura, S; Harada, H; Ohashi, A; Torimura, S

2005-12-01

Treatment of artificial wastewater containing a large amount of suspended solids comprised of soybean processing waste and pig fodder was studied using a novel multi-staged upflow anaerobic sludge blanket reactor. The reactor consisted of three compartments, each containing a gas solid separator. The wastewater had chemical oxygen demand of approximately 21600 mg l(-1), suspended solids of 12800 mg l(-1), and an ammonia concentration of 945 mg l(-1). A continuous experiment without effluent circulation showed that the multi-staged reactor was not that effective for the treatment of wastewater containing a large amount of suspended solids. However, operation of the reactor with circulation of effluent enabled the reactor to achieve organic removal of 85% and approximately 70% methane conversion at loading rates of between 4.0 to 5.4 kg-chemical oxygen demand per cubic meter per day, meaning that the reactor was more effective when effluent was circulated. Morphological investigation revealed that the crude fiber in the sludge was partially degraded and that it had many small depressions on its surface. Evolved biogas may have become caught in these depressions of the fibers and caused washout of the sludge.

75 FR 17786 - Advisory Committee on Reactor Safeguards; Meeting of the ACRS Subcommittee on Power Uprates...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-07

... Boiling Water Reactor Owners Group's (BWROG) topical report NEDC-33347P, ``Containment Overpressure Credit... for the Use of Containment Accident Pressure in Determining the NPSH Margin of ECCS and Containment...

PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blakeman, Edward D; Peplow, Douglas E.; Wagner, John C

2007-09-01

The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally filesmore » and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts.« less

Method of producing pyrolysis gases from carbon-containing materials

DOEpatents

Mudge, Lyle K.; Brown, Michael D.; Wilcox, Wayne A.; Baker, Eddie G.

1989-01-01

A gasification process of improved efficiency is disclosed. A dual bed reactor system is used in which carbon-containing feedstock materials are first treated in a gasification reactor to form pyrolysis gases. The pyrolysis gases are then directed into a catalytic reactor for the destruction of residual tars/oils in the gases. Temperatures are maintained within the catalytic reactor at a level sufficient to crack the tars/oils in the gases, while avoiding thermal breakdown of the catalysts. In order to minimize problems associated with the deposition of carbon-containing materials on the catalysts during cracking, a gaseous oxidizing agent preferably consisting of air, oxygen, steam, and/or mixtures thereof is introduced into the catalytic reactor at a high flow rate in a direction perpendicular to the longitudinal axis of the reactor. This oxidizes any carbon deposits on the catalysts, which would normally cause catalyst deactivation.

LOFT. Reactor arrives at containment building (TAN650), now being pushed ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

LOFT. Reactor arrives at containment building (TAN-650), now being pushed by locomotive. Camera facing northerly. Note "Hello Dolly" and "PWR MTA No. 1" (pressurized water reactor mobile test assembly) signs. Date: 1973. INEEL negative no. 73-3710 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

Biofilm architecture in a novel pressurized biofilm reactor.

PubMed

Jiang, Wei; Xia, Siqing; Duan, Liang; Hermanowicz, Slawomir W

2015-01-01

A novel pure-oxygen pressurized biofilm reactor was operated at different organic loading, mechanical shear and hydrodynamic conditions to understand the relationships between biofilm architecture and its operation. The ultimate goal was to improve the performance of the biofilm reactor. The biofilm was labeled with seven stains and observed with confocal laser scanning microscopy. Unusual biofilm architecture of a ribbon embedded between two surfaces with very few points of attachment was observed. As organic loading increased, the biofilm morphology changed from a moderately rough layer into a locally smoother biomass with significant bulging protuberances, although the chemical oxygen demand (COD) removal efficiency remained unchanged at about 75%. At higher organic loadings, biofilms contained a larger fraction of active cells distributed uniformly within a proteinaceous matrix with decreasing polysaccharide content. Higher hydrodynamic shear in combination with high organic loading resulted in the collapse of biofilm structure and a substantial decrease in reactor performance (a COD removal of 16%). Moreover, the important role of proteins for the spatial distribution of active cells was demonstrated quantitatively.

Functionalizing Microporous Membranes for Protein Purification and Protein Digestion

NASA Astrophysics Data System (ADS)

Dong, Jinlan; Bruening, Merlin L.

2015-07-01

This review examines advances in the functionalization of microporous membranes for protein purification and the development of protease-containing membranes for controlled protein digestion prior to mass spectrometry analysis. Recent studies confirm that membranes are superior to bead-based columns for rapid protein capture, presumably because convective mass transport in membrane pores rapidly brings proteins to binding sites. Modification of porous membranes with functional polymeric films or TiO2 nanoparticles yields materials that selectively capture species ranging from phosphopeptides to His-tagged proteins, and protein-binding capacities often exceed those of commercial beads. Thin membranes also provide a convenient framework for creating enzyme-containing reactors that afford control over residence times. With millisecond residence times, reactors with immobilized proteases limit protein digestion to increase sequence coverage in mass spectrometry analysis and facilitate elucidation of protein structures. This review emphasizes the advantages of membrane-based techniques and concludes with some challenges for their practical application.

Functionalizing Microporous Membranes for Protein Purification and Protein Digestion.

PubMed

Dong, Jinlan; Bruening, Merlin L

2015-01-01

This review examines advances in the functionalization of microporous membranes for protein purification and the development of protease-containing membranes for controlled protein digestion prior to mass spectrometry analysis. Recent studies confirm that membranes are superior to bead-based columns for rapid protein capture, presumably because convective mass transport in membrane pores rapidly brings proteins to binding sites. Modification of porous membranes with functional polymeric films or TiO₂ nanoparticles yields materials that selectively capture species ranging from phosphopeptides to His-tagged proteins, and protein-binding capacities often exceed those of commercial beads. Thin membranes also provide a convenient framework for creating enzyme-containing reactors that afford control over residence times. With millisecond residence times, reactors with immobilized proteases limit protein digestion to increase sequence coverage in mass spectrometry analysis and facilitate elucidation of protein structures. This review emphasizes the advantages of membrane-based techniques and concludes with some challenges for their practical application.

100-kWe lunar/Mars surface power utilizing the SP-100 reactor with dynamic conversion

NASA Technical Reports Server (NTRS)

Harty, Richard B.; Mason, Lee S.

1992-01-01

Results are presented from a study of the coupling of an SP-100 nuclear reactor with either a Stirling or Brayton power system, at the 100 kWe level, for a power generating system suitable for operation in the lunar and Martian surface environments. In the lunar environment, the reactor and primary coolant loop would be contained in a guard vessel to protect from a loss of primary loop containment. For Mars, all refractory components, including the reactor, coolant, and power conversion components will be contained in a vacuum vessel for protection against the CO2 environment.

Tokamak experimental power reactor conceptual design. Volume II

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1976-08-01

Volume II contains the following appendices: (1) summary of EPR design parameters, (2) impurity control, (3) plasma computational models, (4) structural support system, (5) materials considerations for the primary energy conversion system, (6) magnetics, (7) neutronics penetration analysis, (8) first wall stress analysis, (9) enrichment of isotopes of hydrogen by cryogenic distillation, and (10) noncircular plasma considerations. (MOW)

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2011 CFR

2011-01-01

.... Containment inspection. B. Repordkeeping of test results. I. Introduction One of the conditions of all... following: A. Type A test—1. Pretest requirements. (a) Containment inspection in accordance with V. A. shall.... During the period between the completion of one Type A test and the initiation of the containment...

Effect of reactor radiation on the thermal conductivity of TREAT fuel

DOE PAGES

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; ...

2017-02-04

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2more » particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.« less

Effect of reactor radiation on the thermal conductivity of TREAT fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2more » particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.« less

Dense, layered membranes for hydrogen separation

DOEpatents

Roark, Shane E.; MacKay, Richard; Mundschau, Michael V.

2006-02-21

This invention provides hydrogen-permeable membranes for separation of hydrogen from hydrogen-containing gases. The membranes are multi-layer having a central hydrogen-permeable layer with one or more catalyst layers, barrier layers, and/or protective layers. The invention also relates to membrane reactors employing the hydrogen-permeable membranes of the invention and to methods for separation of hydrogen from a hydrogen-containing gas using the membranes and reactors. The reactors of this invention can be combined with additional reactor systems for direct use of the separated hydrogen.

Fluidized bed silicon deposition from silane

NASA Technical Reports Server (NTRS)

Hsu, George C. (Inventor); Levin, Harry (Inventor); Hogle, Richard A. (Inventor); Praturi, Ananda (Inventor); Lutwack, Ralph (Inventor)

1982-01-01

A process and apparatus for thermally decomposing silicon containing gas for deposition on fluidized nucleating silicon seed particles is disclosed. Silicon seed particles are produced in a secondary fluidized reactor by thermal decomposition of a silicon containing gas. The thermally produced silicon seed particles are then introduced into a primary fluidized bed reactor to form a fluidized bed. Silicon containing gas is introduced into the primary reactor where it is thermally decomposed and deposited on the fluidized silicon seed particles. Silicon seed particles having the desired amount of thermally decomposed silicon product thereon are removed from the primary fluidized reactor as ultra pure silicon product. An apparatus for carrying out this process is also disclosed.

Fluidized bed silicon deposition from silane

NASA Technical Reports Server (NTRS)

Hsu, George (Inventor); Levin, Harry (Inventor); Hogle, Richard A. (Inventor); Praturi, Ananda (Inventor); Lutwack, Ralph (Inventor)

1984-01-01

A process and apparatus for thermally decomposing silicon containing gas for deposition on fluidized nucleating silicon seed particles is disclosed. Silicon seed particles are produced in a secondary fluidized reactor by thermal decomposition of a silicon containing gas. The thermally produced silicon seed particles are then introduced into a primary fluidized bed reactor to form a fludized bed. Silicon containing gas is introduced into the primary reactor where it is thermally decomposed and deposited on the fluidized silicon seed particles. Silicon seed particles having the desired amount of thermally decomposed silicon product thereon are removed from the primary fluidized reactor as ultra pure silicon product. An apparatus for carrying out this process is also disclosed.

Radioactive Waste Management and Environmental Contamination Issues at the Chernobyl Site

DOE Office of Scientific and Technical Information (OSTI.GOV)

Napier, Bruce A.; Schmieman, Eric A.; Voitsekhovitch, Oleg V.

2007-11-01

The destruction of the Unit 4 reactor at the Chernobyl Nuclear Power Plant resulted in the generation of radioactive contamination and radioactive waste at the site and in the surrounding area (referred to as the Exclusion Zone). In the course of remediation activities, large volumes of radioactive waste were generated and placed in temporary near surface waste-storage and disposal facilities. Trench and landfill type facilities were created from 1986 to 1987 in the Chernobyl Exclusion Zone at distances 0.5 to 15 km from the NPP site. This large number of facilities was established without proper design documentation, engineered barriers, ormore » hydrogeological investigations and they do not meet contemporary waste-safety requirements. Immediately following the accident, a Shelter was constructed over the destroyed reactor; in addition to uncertainties in stability at the time of its construction, structural elements of the Shelter have degraded as a result of corrosion. The main potential hazard of the Shelter is a possible collapse of its top structures and release of radioactive dust into the environment. A New Safe Confinement (NSC) with a 100-years service life is planned to be built as a cover over the existing Shelter as a longer-term solution. The construction of the NSC will enable the dismantlement of the current Shelter, removal of highly radioactive, fuel-containing materials from Unit 4, and eventual decommissioning of the damaged reactor. More radioactive waste will be generated during NSC construction, possible Shelter dismantling, removal of fuel containing materials, and decommissioning of Unit 4. The future development of the Exclusion Zone depends on the future strategy for converting Unit 4 into an ecologically safe system, i.e., the development of the NSC, the dismantlement of the current Shelter, removal of fuel containing material, and eventual decommissioning of the accident site. To date, a broadly accepted strategy for radioactive waste management at the reactor site and in the Exclusion Zone, and especially for high-level and long-lived waste, has not been developed.« less

Probability of in-vessel steam explosion-induced containment failure for a KWU PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Esmaili, H.; Khatib-Rahbar, M.; Zuchuat, O.

During postulated core meltdown accidents in light water reactors, there is a likelihood for an in-vessel steam explosion when the melt contacts the coolant in the lower plenum. The objective of the work described in this paper is to determine the conditional probability of in-vessel steam explosion-induced containment failure for a Kraftwerk Union (KWU) pressurized water reactor (PWR). The energetics of the explosion depends on the mass of the molten fuel that mixes with the coolant and participates in the explosion and on the conversion of fuel thermal energy into mechanical work. The work can result in the generation ofmore » dynamic pressures that affect the lower head (and possibly lead to its failure), and it can cause acceleration of a slug (fuel and coolant material) upward that can affect the upper internal structures and vessel head and ultimately cause the failure of the upper head. If the upper head missile has sufficient energy, it can reach the containment shell and penetrate it. The analysis, must therefore, take into account all possible dissipation mechanisms.« less

78 FR 53482 - Entergy Operations, Inc., River Bend Station, Unit 1; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-29

... facility consists of a boiling-water reactor located in West Feliciana Parish, Louisiana. 2.0 Request... Containment Leakage Testing for Water- Cooled Power Reactors,'' requires that components which penetrate containment be periodically leak tested at the ``P a, '' defined as the ``calculated peak containment internal...

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

DOEpatents

Travelli, A.

1985-10-25

A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

DOEpatents

Travelli, Armando

1988-01-01

A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

Nuclear reactors built, being built, or planned, 1991

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simpson, B.

1992-07-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor ismore » an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

Safety control circuit for a neutronic reactor

DOEpatents

Ellsworth, Howard C.

2004-04-27

A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, a safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper.

The Air Force Nuclear Engineering Center Structural Activation and Integrity Evaluation

DTIC Science & Technology

1990-03-01

Vi1 List of Figures Figure Page 1. Inside Piqua Nuclear Power Facility containment building on top of the entombed reactor core ... 5...5. Predicted activity percentage of individual materials in the AFNEC ..... ........................ 21 6. Predicted radioisotope activity percentage...of total radioisotopic inventory within entombment at 20 years after shutdown ......................... 23 iv List of Tables Table Page 1. ORIGEN2

Moon base reactor system

NASA Technical Reports Server (NTRS)

Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

1989-01-01

The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tan, Lizhen; Yang, Ying; Sridharan, K.

2015-12-01

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computationalmore » tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe 2+) irradiation.« less

Carbonaceous material for production of hydrogen from low heating value fuel gases

DOEpatents

Koutsoukos, Elias P.

1989-01-01

A process for the catalytic production of hydrogen, from a wide variety of low heating value fuel gases containing carbon monoxide, comprises circulating a carbonaceous material between two reactors--a carbon deposition reactor and a steaming reactor. In the carbon deposition reactor, carbon monoxide is removed from a fuel gas and is deposited on the carbonaceous material as an active carbon. In the steaming reactor, the reactive carbon reacts with steam to give hydrogen and carbon dioxide. The carbonaceous material contains a metal component comprising from about 75% to about 95% cobalt, from about 5% to about 15% iron, and up to about 10% chromium, and is effective in suppressing the production of methane in the steaming reactor.

Catalyst support structure, catalyst including the structure, reactor including a catalyst, and methods of forming same

DOEpatents

Van Norman, Staci A.; Aston, Victoria J.; Weimer, Alan W.

2017-05-09

Structures, catalysts, and reactors suitable for use for a variety of applications, including gas-to-liquid and coal-to-liquid processes and methods of forming the structures, catalysts, and reactors are disclosed. The catalyst material can be deposited onto an inner wall of a microtubular reactor and/or onto porous tungsten support structures using atomic layer deposition techniques.

The effect of the composition of plutonium loaded on the reactivity change and the isotopic composition of fuel produced in a fast reactor

NASA Astrophysics Data System (ADS)

Blandinskiy, V. Yu.

2014-12-01

This paper presents the results of a numerical investigation into burnup and breeding of nuclides in metallic fuel consisting of a mixture of plutonium and depleted uranium in a fast reactor with sodium coolant. The feasibility of using plutonium contained in spent nuclear fuel from domestic thermal reactors and weapons-grade plutonium is discussed. It is shown that the largest production of secondary fuel and the least change in the reactivity over the reactor lifetime can be achieved when employing plutonium contained in spent nuclear fuel from a reactor of the RBMK-1000 type.

Atomic oxygen reactor having at least one sidearm conduit

NASA Technical Reports Server (NTRS)

Koontz, Steven L. (Inventor)

1994-01-01

An apparatus for treating a microporous structure with atomic oxygen is presented. The apparatus includes a main gas chamber for flowing gas in an axial direction and a source of gas, containing atomic oxygen, connected for introducing the gas into the main gas chamber. The apparatus employs at least one side arm extending from the main atomic oxygen-containing chamber. The side arm has characteristic relaxation times such that a uniform atomic oxygen dose rate is delivered to a specimen positioned transversely in the side arm spaced from the main gas chamber.

Gaseous-fuel nuclear reactor research for multimegawatt power in space

NASA Technical Reports Server (NTRS)

Thom, K.; Schneider, R. T.; Helmick, H. H.

1977-01-01

In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

A 1055 ft/sec impact test of a two foot diameter model nuclear reactor containment system without fracture

NASA Technical Reports Server (NTRS)

Puthoff, R. L.

1972-01-01

A study to determine the feasibility of containing the fission products of a mobile reactor in the event of an impact is presented. The model simulated the reactor core, energy absorbing gamma shielding, neutron shielding and the containment vessel. It was impacted against an 18,000 pound reinforced concrete block at 1055 ft/sec. The model was significantly deformed and the concrete block demolished. No leaks were detected nor were any cracks observed in the model after impact.

Nuclear reactor for breeding U.sup.233

DOEpatents

Bohanan, Charles S.; Jones, David H.; Raab, Jr., Harry F.; Radkowsky, Alvin

1976-01-01

A light-water-cooled nuclear reactor capable of breeding U.sup.233 for use in a light-water breeder reactor includes physically separated regions containing U.sup.235 fissile material and U.sup.238 fertile material and Th.sup.232 fertile material and Pu.sup.239 fissile material, if available. Preferably the U.sup.235 fissile material and U.sup.238 fertile material are contained in longitudinally movable seed regions and the Pu.sup.239 fissile material and Th.sup.232 fertile material are contained in blanket regions surrounding the seed regions.

Oxygen transport membrane system and method for transferring heat to catalytic/process reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly, Sean M.; Kromer, Brian R.; Litwin, Michael M.

A method and apparatus for producing heat used in a synthesis gas production process is provided. The disclosed method and apparatus include a plurality of tubular oxygen transport membrane elements adapted to separate oxygen from an oxygen containing stream contacting the retentate side of the membrane elements. The permeated oxygen is combusted with a hydrogen containing synthesis gas stream contacting the permeate side of the tubular oxygen transport membrane elements thereby generating a reaction product stream and radiant heat. The present method and apparatus also includes at least one catalytic reactor containing a catalyst to promote the steam reforming reactionmore » wherein the catalytic reactor is surrounded by the plurality of tubular oxygen transport membrane elements. The view factor between the catalytic reactor and the plurality of tubular oxygen transport membrane elements radiating heat to the catalytic reactor is greater than or equal to 0.5« less

Oxygen transport membrane system and method for transferring heat to catalytic/process reactors

DOEpatents

Kelly, Sean M; Kromer, Brian R; Litwin, Michael M; Rosen, Lee J; Christie, Gervase Maxwell; Wilson, Jamie R; Kosowski, Lawrence W; Robinson, Charles

2014-01-07

A method and apparatus for producing heat used in a synthesis gas production is provided. The disclosed method and apparatus include a plurality of tubular oxygen transport membrane elements adapted to separate oxygen from an oxygen containing stream contacting the retentate side of the membrane elements. The permeated oxygen is combusted with a hydrogen containing synthesis gas stream contacting the permeate side of the tubular oxygen transport membrane elements thereby generating a reaction product stream and radiant heat. The present method and apparatus also includes at least one catalytic reactor containing a catalyst to promote the stream reforming reaction wherein the catalytic reactor is surrounded by the plurality of tubular oxygen transport membrane elements. The view factor between the catalytic reactor and the plurality of tubular oxygen transport membrane elements radiating heat to the catalytic reactor is greater than or equal to 0.5.

The IRIS Spool-Type Reactor Coolant Pump

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kujawski, J.M.; Kitch, D.M.; Conway, L.E.

2002-07-01

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safetymore » to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called 'spool' pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper. (authors)« less

New semi-pilot-scale reactor to study the photocatalytic inactivation of phages contained in aerosol.

PubMed

Briggiler Marcó, Mariángeles; Negro, Antonio Carlos; Alfano, Orlando Mario; Quiberoni, Andrea Del Luján

2017-04-12

The aims of this work were to design and build a photocatalytic reactor (UV-A/TiO 2 ) to study the inactivation of phages contained in bioaerosols, which constitute the main dissemination via phages in industrial environments. The reactor is a close system with recirculation that consists of a stainless steel camera (cubic form, side of 60 cm) in which air containing the phage particles circulates and an acrylic compartment with six borosilicate plates covered with TiO 2 . The reactor is externally illuminated by 20 UV-A lamps. Both compartments are connected by a fan to facilitate the sample circulation. Samples are injected into the camera using two piston nebulizers working in series whereas several methodologies for sampling (impinger/syringe, sampling on photocatalytic plates, and impact of air on slide) were assayed. The reactor setup was carried out using phage B1 (Lactobacillus plantarum), and assays demonstrated a decrease of phage counts of 2.7 log orders after 1 h of photocatalytic treatment. Photonic efficiencies of inactivation were assessed by phage sampling on the photocatalytic plates or by impact of air on a glass slide at the photocatalytic reactor exit. Efficiencies of the same order of magnitude were observed using both sampling methods. This study demonstrated that the designed photocatalytic reactor is effective to inactivate phage B1 (Lb. plantarum) contained in bioaerosols.

Method and apparatus for chemically altering fluids in continuous flow

DOEpatents

Heath, W.O.; Virden, J.W. Jr.; Richardson, R.L.; Bergsman, T.M.

1993-10-19

The present invention relates to a continuous flow fluid reactor for chemically altering fluids. The reactor operates on standard frequency (50 to 60 Hz) electricity. The fluid reactor contains particles that are energized by the electricity to form a corona throughout the volume of the reactor and subsequently a non-equilibrium plasma that interacts with the fluid. Particles may form a fixed bed or a fluid bed. Electricity may be provided through electrodes or through an inductive coil. Fluids include gases containing exhaust products and organic fuels requiring oxidation. 4 figures.

Method and apparatus for chemically altering fluids in continuous flow

DOEpatents

Heath, William O.; Virden, Jr., Judson W.; Richardson, R. L.; Bergsman, Theresa M.

1993-01-01

The present invention relates to a continuous flow fluid reactor for chemically altering fluids. The reactor operates on standard frequency (50 to 60 Hz) electricity. The fluid reactor contains particles that are energized by the electricity to form a corona throughout the volume of the reactor and subsequently a non-equilibrium plasma that interacts with the fluid. Particles may form a fixed bed or a fluid bed. Electricity may be provided through electrodes or through an inductive coil. Fluids include gases containing exhaust products and organic fuels requiring oxidation.

CFD Analyses of Air-Ingress Accident for VHTRs

NASA Astrophysics Data System (ADS)

Ham, Tae Kyu

The Very High Temperature Reactor (VHTR) is one of six proposed Generation-IV concepts for the next generation of nuclear powered plants. The VHTR is advantageous because it is able to operate at very high temperatures, thus producing highly efficient electrical generation and hydrogen production. A critical safety event of the VHTR is a loss-of-coolant accident. This accident is initiated, in its worst-case scenario, by a double-ended guillotine break of the cross vessel that connects the reactor vessel and the power conversion unit. Following the depressurization process, the air (i.e., the air and helium mixture) in the reactor cavity could enter the reactor core causing an air-ingress event. In the event of air-ingress into the reactor core, the high-temperature in-core graphite structures will chemically react with the air and could lose their structural integrity. We designed a 1/8th scaled-down test facility to develop an experimental database for studying the mechanisms involved in the air-ingress phenomenon. The current research focuses on the analysis of the air-ingress phenomenon using the computational fluid dynamics (CFD) tool ANSYS FLUENT for better understanding of the air-ingress phenomenon. The anticipated key steps in the air-ingress scenario for guillotine break of VHTR cross vessel are: 1) depressurization; 2) density-driven stratified flow; 3) local hot plenum natural circulation; 4) diffusion into the reactor core; and 5) global natural circulation. However, the OSU air-ingress test facility covers the time from depressurization to local hot plenum natural circulation. Prior to beginning the CFD simulations for the OSU air-ingress test facility, benchmark studies for the mechanisms which are related to the air-ingress accident, were performed to decide the appropriate physical models for the accident analysis. In addition, preliminary experiments were performed with a simplified 1/30th scaled down acrylic set-up to understand the air-ingress mechanism and to utilize the CFD simulation in the analysis of the phenomenon. Previous air-ingress studies simulated the depressurization process using simple assumptions or 1-D system code results. However, recent studies found flow oscillations near the end of the depressurization which could influence the next stage of the air-ingress accident. Therefore, CFD simulations were performed to examine the air-ingress mechanisms from the depressurization through the establishment of local natural circulation initiate. In addition to the double-guillotine break scenario, there are other scenarios that can lead to an air-ingress event such as a partial break were in the cross vessel with various break locations, orientations, and shapes. These additional situations were also investigated. The simulation results for the OSU test facility showed that the discharged helium coolant from a reactor vessel during the depressurization process will be mixed with the air in the containment. This process makes the density of the gas mixture in the containment lower and the density-driven air-ingress flow slower because the density-driven flow is established by the density difference of the gas species between the reactor vessel and the containment. In addition, for the simulations with various initial and boundary conditions, the simulation results showed that the total accumulated air in the containment collapsed within 10% standard deviation by: 1. multiplying the density ratio and viscosity ratio of the gas species between the containment and the reactor vessel and 2. multiplying the ratio of the air mole fraction and gas temperature to the reference value. By replacing the gas mixture in the reactor cavity with a gas heavier than the air, the air-ingress speed slowed down. Based on the understanding of the air-ingress phenomena for the GT-MHR air-ingress scenario, several mitigation measures of air-ingress accident are proposed. The CFD results are utilized to plan experimental strategy and apparatus installation to obtain the best results when conducting an experiment. The validation of the generated CFD solutions will be performed with the OSU air-ingress experimental results. (Abstract shortened by UMI.).

Mechanical characteristics of heterogeneous structures obtained by high-temperature brazing of corrosion-resistant steels with rapidly quenched non-boron nickel-based alloys

NASA Astrophysics Data System (ADS)

Kalin, B.; Penyaz, M.; Ivannikov, A.; Sevryukov, O.; Bachurina, D.; Fedotov, I.; Voennov, A.; Abramov, E.

2018-01-01

Recently, the use rapidly quenched boron-containing nickel filler metals for high temperature brazing corrosion resistance steels different classes is perspective. The use of these alloys leads to the formation of a complex heterogeneous structure in the diffusion zone that contains separations of intermediate phases such as silicides and borides. This structure negatively affects the strength characteristics of the joint, especially under dynamic loads and in corrosive environment. The use of non-boron filler metals based on the Ni-Si-Be system is proposed to eliminate this structure in the brazed seam. Widely used austenitic 12Cr18Ni10Ti and ferrite-martensitic 16Cr12MoSiWNiVNb reactor steels were selected for research and brazing was carried out. The mechanical characteristics of brazed joints were determined using uniaxial tensile and impact toughness tests, and fractography was investigated by electron microscopy.

Evaluation of Maximum Radionuclide Groundwater Concentrations for Basement Fill Model. Zion Station Restoration Project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sullivan, T.

2016-05-20

ZionSolutions is in the process of decommissioning the Zion Nuclear Power Station (ZNPS). After decommissioning is completed, the site will contain two reactor Containment Buildings, the Fuel Handling Building and Transfer Canals, Auxiliary Building, Turbine Building, Crib House/Forebay, and a Waste Water Treatment Facility that have been demolished to a depth of 3 feet below grade. Additional below ground structures remaining will include the Main Steam Tunnels and large diameter intake and discharge pipes. These additional structures are not included in the modeling described in this report, but the inventory remaining (expected to be very low) will be included withmore » one of the structures that are modeled as designated in the Zion Station Restoration Project (ZSRP) License Termination Plan (LTP). The remaining underground structures will be backfilled with clean material. The final selection of fill material has not been made.« less

Nitrate treatment effects on bacterial community biofilm formed on carbon steel in produced water stirred tank bioreactor.

PubMed

Marques, Joana Montezano; de Almeida, Fernando Pereira; Lins, Ulysses; Seldin, Lucy; Korenblum, Elisa

2012-06-01

To better understand the impact of nitrate in Brazilian oil reservoirs under souring processes and corrosion, the goal of this study was to analyse the effect of nitrate on bacterial biofilms formed on carbon steel coupons using reactors containing produced water from a Brazilian oil platform. Three independent experiments were carried out (E1, E2 and E3) using the same experimental conditions and different incubation times (5, 45 and 80 days, respectively). In every experiment, two biofilm-reactors were operated: one was treated with continuous nitrate flow (N reactor), and the other was a control reactor without nitrate (C reactor). A Polymerase Chain Reaction-Denaturing Gradient Gel Electrophoresis approach using the 16S rRNA gene was performed to compare the bacterial groups involved in biofilm formation in the N and C reactors. DGGE profiles showed remarkable changes in community structure only in experiments E2 and E3. Five bands extracted from the gel that represented the predominant bacterial groups were identified as Bacillus aquimaris, B. licheniformis, Marinobacter sp., Stenotrophomonas maltophilia and Thioclava sp. A reduction in the sulfate-reducing bacteria (SRB) most probable number counts was observed only during the longer nitrate treatment (E3). Carbon steel coupons used for biofilm formation had a slightly higher weight loss in N reactors in all experiments. When the coupon surfaces were analysed by scanning electron microscopy, an increase in corrosion was observed in the N reactors compared with the C reactors. In conclusion, nitrate reduced the viable SRB counts. Nevertheless, the nitrate dosing increased the pitting of coupons.

Study on ( n,t) Reactions of Zr, Nb and Ta Nuclei

NASA Astrophysics Data System (ADS)

Tel, E.; Yiğit, M.; Tanır, G.

2012-04-01

The world faces serious energy shortages in the near future. To meet the world energy demand, the nuclear fusion with safety, environmentally acceptability and economic is the best suited. Fusion is attractive as an energy source because of the virtually inexhaustible supply of fuel, the promise of minimal adverse environmental impact, and its inherent safety. Fusion will not produce CO2 or SO2 and thus will not contribute to global warming or acid rain. Furthermore, there are not radioactive nuclear waste problems in the fusion reactors. Although there have been significant research and development studies on the inertial and magnetic fusion reactor technology, there is still a long way to go to penetrate commercial fusion reactors to the energy market. Because, tritium self-sufficiency must be maintained for a commercial power plant. For self-sustaining (D-T) fusion driver tritium breeding ratio should be greater than 1.05. And also, the success of fusion power system is dependent on performance of the first wall, blanket or divertor systems. So, the performance of structural materials for fusion power systems, understanding nuclear properties systematic and working out of ( n,t) reaction cross sections are very important. Zirconium (Zr), Niobium (Nb) and Tantal (Ta) containing alloys are important structural materials for fusion reactors, accelerator-driven systems, and many other fields. In this study, ( n,t) reactions for some structural fusion materials such as 88,90,92,94,96Zr, 93,94,95Nb and 179,181Ta have been investigated. The calculated results are discussed andcompared with the experimental data taken from the literature.

75 FR 21046 - Advisory Committee on Reactor Safeguards

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-22

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards In accordance with the... on Reactor Safeguards (ACRS) will hold a meeting on May 6-8, 2010, 11545 Rockville Pike, Rockville....: Boiling Water Reactor (BWR) Owners Group (BWROG) Topical Report NEDC-33347P, ``Containment Overpressure...

Self-actuated shutdown system for a commercial size LMFBR. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dupen, C.F.G.

1978-08-01

A Self-Actuated Shutdown System (SASS) is defined as a reactor shutdown system in which sensors, release mechanisms and neutron absorbers are contained entirely within the reactor core structure, where they respond inherently to abnormal local process conditions, by shutting down the reactor, independently of the plant protection system (PPS). It is argued that a SASS, having a response time similar to that of the PPS, would so reduce the already very low probability of a failure-to-scram event that costly design features, derived from core disruptive accident analysis, could be eliminated. However, the thrust of the report is the feasibility andmore » reliability of the in-core SASS hardware to achieve sufficiently rapid shutdown. A number of transient overpower and transient undercooling-responsive systems were investigated leading to the selection of a primary candidate and a backup concept. During a transient undercooling event, the recommended device is triggered by the associated rate of change of pressure, whereas the alternate concept responds to the reduction in core pressure drop and requires calibration and adjustment by the operators to accommodate changes in reactor power.« less

Lessons Learned about Liquid Metal Reactors from FFTF Experience

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.

2016-09-20

The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens.more » In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports« less

The Angra Project: Monitoring Nuclear Reactors with Antineutrino Detectors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anjos, J. C.; Barbosa, A. F.; Lima, H. P. Jr.

2010-03-30

We present the status of the Angra Neutrino project, describing the development of an antineutrino detector aimed at monitoring nuclear reactor activity. The experiment will take place at the Brazilian nuclear power plant located in Angra dos Reis. The Angra II reactor, with 4 GW of thermal power, will be used as a source of antineutrinos. A water Cherenkov detector will be placed above ground in a commercial container outside the reactor containment, about 30 m from the reactor core. With a detector of one ton scale a few thousand antineutrino interactions per day are expected. We intend, in amore » first step, to use the measured neutrino event rate to monitor the on--off status and the thermal power delivered by the reactor. In addition to the safeguards issues the project will provide an alternative tool to have an independent measurement of the reactor power.« less

The Angra Project: Monitoring Nuclear Reactors with Antineutrino Detectors

NASA Astrophysics Data System (ADS)

Anjos, J. C.; Barbosa, A. F.; Bezerra, T. J. C.; Chimenti, P.; Gonzalez, L. F. G.; Kemp, E.; de Oliveira, M. A. Leigui; Lima, H. P.; Lima, R. M.; Nunokawa, H.

2010-03-01

We present the status of the Angra Neutrino project, describing the development of an antineutrino detector aimed at monitoring nuclear reactor activity. The experiment will take place at the Brazilian nuclear power plant located in Angra dos Reis. The Angra II reactor, with 4 GW of thermal power, will be used as a source of antineutrinos. A water Cherenkov detector will be placed above ground in a commercial container outside the reactor containment, about 30 m from the reactor core. With a detector of one ton scale a few thousand antineutrino interactions per day are expected. We intend, in a first step, to use the measured neutrino event rate to monitor the on—off status and the thermal power delivered by the reactor. In addition to the safeguards issues the project will provide an alternative tool to have an independent measurement of the reactor power.

Operator Support System Design forthe Operation of RSG-GAS Research Reactor

NASA Astrophysics Data System (ADS)

Santoso, S.; Situmorang, J.; Bakhri, S.; Subekti, M.; Sunaryo, G. R.

2018-02-01

The components of RSG-GAS main control room are facing the problem of material ageing and technology obsolescence as well, and therefore the need for modernization and refurbishment are essential. The modernization in control room can be applied on the operator support system which bears the function in providing information for assisting the operator in conducting diagnosis and actions. The research purpose is to design an operator support system for RSG-GAS control room. The design was developed based on the operator requirement in conducting task operation scenarios and the reactor operation characteristics. These scenarios include power operation, low power operation and shutdown/scram reactor. The operator support system design is presented in a single computer display which contains structure and support system elements e.g. operation procedure, status of safety related components and operational requirements, operation limit condition of parameters, alarm information, and prognosis function. The prototype was developed using LabView software and consisted of components structure and features of the operator support system. Information of each component in the operator support system need to be completed before it can be applied and integrated in the RSG-GAS main control room.

Ukraine: Current Issues and U.S. Policy

DTIC Science & Technology

2012-05-10

the Chernobyl nuclear reactor site. In total, the United States has contributed almost $240 million to Chernobyl cleanup effort. At an international...pledging conference for Chernobyl in April 2011, the United States pledged another $123 million.9 The two countries are cooperating on other...dedicated to improving the safety of the Chernobyl nuclear facility, including finishing the construction of the containment structure over the damaged

HEAT TREATED U-Nb ALLOYS

DOEpatents

McGeary, R.K.; Justusson, W.M.

1959-11-24

A fuel element for a nuclear reactor is described comprising an alloy containing uranium and from 7 to 20 wt.% niobium, the alloy being substantially in the gamma phase and having been produced by working an ingot of the alloy into the desired shape, homogenizing it by annealing it at a temperature in the gamma phase field, and quenching it to retain the gamma phase structure of the alloy.

Impact of conversion to mixed-oxide fuels on reactor structural components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yahr, G.T.

1997-04-01

The use of mixed-oxide (MOX) fuel to replace conventional uranium fuel in commercial light-water power reactors will result in an increase in the neutron flux. The impact of the higher flux on the structural integrity of reactor structural components must be evaluated. This report briefly reviews the effects of radiation on the mechanical properties of metals. Aging degradation studies and reactor operating experience provide a basis for determining the areas where conversion to MOX fuels has the potential to impact the structural integrity of reactor components.

A 640 foot per second impact test of a two foot diameter model nuclear reactor containment system without fracture

NASA Technical Reports Server (NTRS)

Puthoff, R. L.

1971-01-01

An impact test was conducted on an 1142 pound 2 foot diameter sphere model. The purpose of this test was to determine the feasibility of containing the fission products of a mobile reactor in an impact. The model simulated the reactor core, energy absorbing gamma shielding, neutron shielding and the containment vessel. It was impacted against an 18,000 pound reinforced concrete block. The model was significantly deformed and the concrete block demolished. No leaks were detected nor cracks observed in the model after impact.

Contamination of the transformer oil of power transformers and shunting reactors by metal-containing colloidal particles

DOE Office of Scientific and Technical Information (OSTI.GOV)

L'vov, S. Yu.; Komarov, V. B.; Bondareva, V. N.

The results of a measurement of the contamination of the oil in 66 transformers by metal-containing colloidal particles, formed as a result of the interaction of the oil with the structural materials (the copper of the windings, the iron of the tank and core etc.), and also the results of measurements of the optical turbidity of the oil in 136 transformers when they were examined at the Power Engineering Research and Development Center Company are presented. Methods of determining the concentration of copper and iron in transformer oil are considered. The limiting values of the optical turbidity factors, the coppermore » and iron content are determined. These can serve as a basis for taking decisions on whether to replace the silica gel of the filters for continuously purifying the oil of power transformers and the shunting reactors in addition to the standardized oil contamination factors, namely, the dielectric loss tangent and the acidity number of the oil.« less

Reactor Application for Coaching Newbies

DOE Office of Scientific and Technical Information (OSTI.GOV)

2015-06-17

RACCOON is a Moose based reactor physics application designed to engage undergraduate and first-year graduate students. The code contains capabilities to solve the multi group Neutron Diffusion equation in eigenvalue and fixed source form and will soon have a provision to provide simple thermal feedback. These capabilities are sufficient to solve example problems found in Duderstadt & Hamilton (the typical textbook of senior level reactor physics classes). RACCOON does not contain any advanced capabilities as found in YAK.

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.

This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARCmore » and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.« less

NUCLEAR REACTOR

DOEpatents

Treshow, M.

1961-09-01

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

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear... requirements for immediate notification of the NRC by licensed operating nuclear power reactors are contained...

Performance assessment of conventional and base-isolated nuclear power plants for earthquake and blast loadings

NASA Astrophysics Data System (ADS)

Huang, Yin-Nan

Nuclear power plants (NPPs) and spent nuclear fuel (SNF) are required by code and regulations to be designed for a family of extreme events, including very rare earthquake shaking, loss of coolant accidents, and tornado-borne missile impacts. Blast loading due to malevolent attack became a design consideration for NPPs and SNF after the terrorist attacks of September 11, 2001. The studies presented in this dissertation assess the performance of sample conventional and base isolated NPP reactor buildings subjected to seismic effects and blast loadings. The response of the sample reactor building to tornado-borne missile impacts and internal events (e.g., loss of coolant accidents) will not change if the building is base isolated and so these hazards were not considered. The sample NPP reactor building studied in this dissertation is composed of containment and internal structures with a total weight of approximately 75,000 tons. Four configurations of the reactor building are studied, including one conventional fixed-base reactor building and three base-isolated reactor buildings using Friction Pendulum(TM), lead rubber and low damping rubber bearings. The seismic assessment of the sample reactor building is performed using a new procedure proposed in this dissertation that builds on the methodology presented in the draft ATC-58 Guidelines and the widely used Zion method, which uses fragility curves defined in terms of ground-motion parameters for NPP seismic probabilistic risk assessment. The new procedure improves the Zion method by using fragility curves that are defined in terms of structural response parameters since damage and failure of NPP components are more closely tied to structural response parameters than to ground motion parameters. Alternate ground motion scaling methods are studied to help establish an optimal procedure for scaling ground motions for the purpose of seismic performance assessment. The proposed performance assessment procedure is used to evaluate the vulnerability of the conventional and base-isolated NPP reactor buildings. The seismic performance assessment confirms the utility of seismic isolation at reducing spectral demands on secondary systems. Procedures to reduce the construction cost of secondary systems in isolated reactor buildings are presented. A blast assessment of the sample reactor building is performed for an assumed threat of 2000 kg of TNT explosive detonated on the surface with a closest distance to the reactor building of 10 m. The air and ground shock waves produced by the design threat are generated and used for performance assessment. The air blast loading to the sample reactor building is computed using a Computational Fluid Dynamics code Air3D and the ground shock time series is generated using an attenuation model for soil/rock response. Response-history analysis of the sample conventional and base isolated reactor buildings to external blast loadings is performed using the hydrocode LS-DYNA. The spectral demands on the secondary systems in the isolated reactor building due to air blast loading are greater than those for the conventional reactor building but much smaller than those spectral demands associated with Safe Shutdown Earthquake shaking. The isolators are extremely effective at filtering out high acceleration, high frequency ground shock loading.

Development of the cascade inertial-confinement-fusion reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pitts, J.H.

Caqscade, originally conceived as a football-shaped, steel-walled reactor containing a Li/sub 2/O granule blanket, is now envisaged as a double-cone-shaped reactor containing a two-layered (three-zone) flowing blanket of BeO and LiAlO/sub 2/ granules. Average blanket exit temperature is 1670 K and gross plant efficiency (net thermal conversion efficiency) using a Brayton cycle is 55%. The reactor has a low-activation SiC-tiled wall. It rotates at 50 rpm, and the granules are transported to the top of the heat exchanger using their peripheral speed; no conveyors or lifts are required. The granules return to the reactor by gravity. After considerable analysis andmore » experimentation, we continue to regard Cascade as a promising reactor concept with the advantages of safety, efficiency, and low activation.« less

Development of the cascade inertial-confinement-fusion reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pitts, J.H.

Cascade, originally conceived as a football-shaped, steel-walled reactor containing a Li/sub 2/O granule blanket, is now envisaged as a double-cone-shaped reactor containing a two-layered (three-zone) flowing blanket of BeO and LiAlO/sub 2/ granules. Average blanket exit temperature is 1670/sup 0/K and gross plant efficiency (net thermal conversion efficiency) using a Brayton cycle is 55%. The reactor has a low-activation SiC-tiled wall. It rotates at 50 rpm, and the granules are transported to the top of the heat exchanger using their peripheral speed; no conveyors or lifts are required. The granules return to the reactor by gravity. After considerable analysis andmore » experimentation, we continue to regard Cascade as a promising reactor concept with the advantages of safety, efficiency, and low activation.« less

Two-stage dehydration of sugars

DOEpatents

Holladay, Johnathan E [Kennewick, WA; Hu, Jianli [Kennewick, WA; Wang, Yong [Richland, WA; Werpy, Todd A [West Richland, WA

2009-11-10

The invention includes methods for producing dianhydrosugar alcohol by providing an acid catalyst within a reactor and passing a starting material through the reactor at a first temperature. At least a portion of the staring material is converted to a monoanhydrosugar isomer during the passing through the column. The monoanhydrosugar is subjected to a second temperature which is greater than the first to produce a dianhydrosugar. The invention includes a method of producing isosorbide. An initial feed stream containing sorbitol is fed into a continuous reactor containing an acid catalyst at a temperature of less than 120.degree. C. The residence time for the reactor is less than or equal to about 30 minutes. Sorbitol converted to 1,4-sorbitan in the continuous reactor is subsequently provided to a second reactor and is dehydrated at a temperature of at least 120.degree. C. to produce isosorbide.

Containment system for supercritical water oxidation reactor

DOEpatents

Chastagner, Philippe

1994-01-01

A system for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary.

Containment system for supercritical water oxidation reactor

DOEpatents

Chastagner, P.

1994-07-05

A system is described for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary. 2 figures.

78 FR 35990 - All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-14

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0128] All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos. (As Shown In Attachment 1), License Nos. (As Shown In Attachment 1), EA-13-109; Order Modifying Licenses With Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident...

Superconducting RF Linacs Driving Subcritical Reactors for Profitable Disposition of Surplus Weapons-grade Plutonium

NASA Astrophysics Data System (ADS)

Cummings, Mary Anne; Johnson, Rolland

Acceptable capital and operating costs of high-power proton accelerators suitable for profitable commercial electric-power and process-heat applications have been demonstrated. However, studies have pointed out that even a few hundred trips of an accelerator lasting a few seconds would lead to unacceptable thermal stresses as each trip causes fission to be turned off in solid fuel structures found in conventional reactors. The newest designs based on the GEM*STAR concept take such trips in stride by using molten-salt fuel, where fuel pin fatigue is not an issue. Other aspects of the GEM*STAR concept which address all historical reactor failures include an internal spallation neutron target and high temperature molten salt fuel with continuous purging of volatile radioactive fission products such that the reactor contains less than a critical mass and almost a million times fewer volatile radioactive fission products than conventional reactors. GEM*STAR is a reactor that without redesign will burn spent nuclear fuel, natural uranium, thorium, or surplus weapons material. It will operate without the need for a critical core, fuel enrichment, or reprocessing making it an excellent candidate for export. As a first application, the design for a pilot plant is described for the profitable disposition of surplus weapons-grade plutonium by using process heat to produce green diesel fuel for the Department of Defense (DOD) from natural gas and renewable carbon.

77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-12

... for Passive Advanced Light Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... Passive Advanced Light Water Reactors.'' The current SRP does not contain guidance on the proposed RTNSS for Passive Advance Light Water Reactors. DATES: Submit comments by November 13, 2012. Comments...

A Semi-Batch Reactor Experiment for the Undergraduate Laboratory

ERIC Educational Resources Information Center

Derevjanik, Mario; Badri, Solmaz; Barat, Robert

2011-01-01

This experiment and analysis offer an economic yet challenging semi-batch reactor experience. Household bleach is pumped at a controlled rate into a batch reactor containing pharmaceutical hydrogen peroxide solution. Batch temperature, product molecular oxygen, and the overall change in solution conductivity are metered. The reactor simulation…

REACTOR FUEL ELEMENTS TESTING CONTAINER

DOEpatents

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

1963-01-15

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

LOFT. Reactor support apparatus inside containment building (TAN650). Camera is ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

LOFT. Reactor support apparatus inside containment building (TAN-650). Camera is on crane rail level and facing northerly. View shows top two banks of round conduit openings on wall for electrical and other connections to control room. Ladders and platforms provide access to reactor instrumentation. Note hatch in floor and drain at edge of floor near wall. Date: 1974. INEEL negative no. 74-219 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

Improvement of yields and rates during enzymatic hydrolysis of cellulose to glucose. Progress report, March 1, 1979-May 31, 1979

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sundstrom, D W; Klei, H E; Coughlin, R W

1979-05-01

The objective of this program is to show that the conversion of cellulose to glucose can be significantly increased by enzymatically removing the inhibitory cellobiose from the reaction system using immobilized ..beta..-glucosidase (..beta..-G). An enzymatic catalyst was prepared and used in a fluidized bed with cellobiose as the substrate, only a 10% loss of activity was observed during a 500 hour period. Cellulose was hydrolyzed in two batch reactors operated side-by-side, with one reactor containing immobilized ..beta..-G and cellulose and the other reactor containing an equal amount of cellulose only. After 30 hours the reactor containing the immobilized ..beta..-G hadmore » 100% more glucose, indicating that the catalytic removal of the cellobiose had a significant effect upon the production of glucose.« less

Deposition of RuO 4 on various surfaces in a nuclear reactor containment

NASA Astrophysics Data System (ADS)

Holm, Joachim; Glänneskog, Henrik; Ekberg, Christian

2009-07-01

During a severe nuclear reactor accident with air ingress, ruthenium can be released from the nuclear fuel in the form of ruthenium tetroxide. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. The aim of this work was to investigate the deposition of gaseous ruthenium tetroxide on aluminium, copper and zinc, which all appear in relatively large amounts in reactor containment. The experiments show that ruthenium tetroxide is deposited on all the metal surfaces, especially on the copper and zinc surfaces. A large deposition of ruthenium tetroxide also appeared on the relatively inert glass surfaces in the experimental set-ups. The analyses of the different surfaces, with several analytical methods, showed that the form of deposited ruthenium was mainly ruthenium dioxide.

Methanation assembly using multiple reactors

DOEpatents

Jahnke, Fred C.; Parab, Sanjay C.

2007-07-24

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

Nuclear power industry: Tendencies in the world and Ukraine

NASA Astrophysics Data System (ADS)

Babenko, V. A.; Jenkovszky, L. L.; Pavlovych, V. N.

2007-11-01

This review deals with new trends in nuclear reactors physics. It opens by an easily understood introduction to nuclear fission energy physics, starting with some history, including the achievements of the Kharkov nuclear physics school. Attention has been given to the development of fission theory, the Strutinsky theory, and the possible use of "nonstandard" fissile elements. The evolution of the design of nuclear reactors, including the merits and demerits of various structures used worldwide, is given in detail. A detailed description of nuclear power plants operating in Ukraine and their (large!) contribution to Ukraine's total electricity production as compared with other countries is presented. A comparative evaluation of different energy sources influencing environment contamination and the pollution caused by the Chernobyl accident are presented. The lessons of the Chernobyl accident are summarized, including the features of the shelter ("Sarkofag") covering the remaining of the power plant fourth block and some examples of calculations of the radioactive evolution of the station's fuel-containing mass (by authors of the present review). The evolution of traditional nuclear reactors designs set forth under the separate heading of next-generation reactors including new projects such as subcritical assemblies controlled by an external beam of particles (neutrons and protons). The Feoktistov reactor operation and the possibility of its realization are discussed among the new ideas.

MODULAR CORE UNITS FOR A NEUTRONIC REACTOR

DOEpatents

Gage, J.F. Jr.; Sherer, D.B.

1964-04-01

A modular core unit for use in a nuclear reactor is described. Many identical core modules can be placed next to each other to make up a complete core. Such a module includes a cylinder of moderator material surrounding a fuel- containing re-entrant coolant channel. The re-entrant channel provides for the circulation of coolant such as liquid sodium from one end of the core unit, through the fuel region, and back out through the same end as it entered. Thermal insulation surrounds the moderator exterior wall inducing heat to travel inwardly to the coolant channel. Spaces between units may be used to accommodate control rods and support structure, which may be cooled by a secondary gas coolant, independently of the main coolant. (AEC)

Posttest analysis of a 1:6-scale reinforced concrete reactor containment building

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weatherby, J.R.

In an experiment conducted at Sandia National Laboratories, 1:6-scale model of a reinforced concrete light water reactor containment building was pressurized with nitrogen gas to more than three times its design pressure. The pressurization produced one large tear and several smaller tears in the steel liner plate that functioned as the primary pneumatic seal for the structure. The data collected from the overpressurization test have been used to evaluate and further refine methods of structural analysis that can be used to predict the performance of containment buildings under conditions produced by a severe accident. This report describes posttest finite elementmore » analyses of the 1:6-scale model tests and compares pretest predictions of the structural response to the experimental results. Strain and displacements calculated in axisymmetric finite element analyses of the 1:6-scale model are compared to strains and displacement measured in the experiment. Detailed analyses of the liner plate are also described in the report. The region of the liner surrounding the large tear was analyzed using two different two-dimensional finite elements model. The results from these analyzed indicate that the primary mechanisms that initiated the tear can be captured in a two- dimensional finite element model. Furthermore, the analyses show that studs used to anchor the liner to the concrete wall, played an important role in initiating the liner tear. Three-dimensional finite element analyses of liner plates loaded by studs are also presented. Results from the three-dimensional analyses are compared to results from two-dimensional analyses of the same problems. 12 refs., 56 figs., 1 tab.« less

Apparatus and method for simulating material damage from a fusion reactor

DOEpatents

Smith, Dale L.; Greenwood, Lawrence R.; Loomis, Benny A.

1989-01-01

An apparatus and method for simulating a fusion environment on a first wall or blanket structure. A material test specimen is contained in a capsule made of a material having a low hydrogen solubility and permeability. The capsule is partially filled with a lithium solution, such that the test specimen is encapsulated by the lithium. The capsule is irradiated by a fast fission neutron source.

Apparatus and method for simulating material damage from a fusion reactor

DOEpatents

Smith, D.L.; Greenwood, L.R.; Loomis, B.A.

1988-05-20

This paper discusses an apparatus and method for simulating a fusion environment on a first wall or blanket structure. A material test specimen is contained in a capsule made of a material having a low hydrogen solubility and permeability. The capsule is partially filled with a lithium solution, such that the test specimen is encapsulated by the lithium. The capsule is irradiated by a fast fission neutron source.

Apparatus and method for simulating material damage from a fusion reactor

DOEpatents

Smith, Dale L.; Greenwood, Lawrence R.; Loomis, Benny A.

1989-03-07

An apparatus and method for simulating a fusion environment on a first wall or blanket structure. A material test specimen is contained in a capsule made of a material having a low hydrogen solubility and permeability. The capsule is partially filled with a lithium solution, such that the test specimen is encapsulated by the lithium. The capsule is irradiated by a fast fission neutron source.

Neutronic Reactor Design to Reduce Neutron Loss

DOEpatents

Miles, F. T.

1961-05-01

A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall. The wall is surrounded by successive layers of pure fertile material and moderator containing fertile material. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. Since the steel has a smaller capture cross section for the fast neutrons, greater nunnbers of neutrons will pass into the blanket, thereby increasing the over-all efficiency of the reactor. (AEC)

Dismantling the nuclear research reactor Thetis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michiels, P.

The research reactor Thetis, in service since 1967 and stopped in 2003, is part of the laboratories of the institution of nuclear science of the University of Ghent. The reactor, of the pool-type, was used as a neutron-source for the production of radio-isotopes and for activation analyses. The reactor is situated in a water pool with inner diameter of 3 m. and a depth of 7.5 m. The reactor core is situated 5.3 m under water level. Besides the reactor, the pool contains pneumatic loops, handling tools, graphite blocks for neutron moderation and other experimental equipment. The building houses storagemore » rooms for fissile material and sources, a pneumatic circuit for transportation of samples, primary and secondary cooling circuits, water cleaning resin circuits, a ventilation system and other necessary devices. Because of the experimental character of the reactor, laboratories with glove boxes and other tools were needed and are included in the dismantling program. The building is in 3 levels with a crawl-space. The ground-floor contains the ventilation installation, the purification circuits with tanks, cooling circuits and pneumatic transport system. On the first floor, around the reactor hall, the control-room, visiting area, end-station for pneumatic transport, waste-storage room, fuel storage room and the labs are located. The second floor contains a few laboratories and end stations of the two high speed transfer tubes. The lowest level of the pool is situated under ground level. The reactor has been operated at a power of 150 kW and had a max operating power of 250 kW. Belgoprocess has been selected to decommission the reactor, the labs, storage halls and associated circuits to free release the building for conventional reuse and for the removal of all its internals as legal defined. Besides the dose-rate risk and contamination risk, there is also an asbestos risk of contamination. During construction of the installation, asbestos-containing materials were used, which must be removed in controlled conditions. The ventilation system is considered free from nuclear contamination but it contains asbestos. This paper covers the organization of the dismantling work, the technical execution aspect and conclusions already known (dismantling is ongoing as this is written). (authors)« less

RADIATION FACILITY FOR NUCLEAR REACTORS

DOEpatents

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

1961-12-12

A radiation facility is designed for irradiating samples in close proximity to the core of a nuclear reactor. The facility comprises essentially a tubular member extending through the biological shield of the reactor and containing a manipulatable rod having the sample carrier at its inner end, the carrier being longitudinally movable from a position in close proximity to the reactor core to a position between the inner and outer faces of the shield. Shield plugs are provided within the tubular member to prevent direct radiation from the core emanating therethrough. In this device, samples may be inserted or removed during normal operation of the reactor without exposing personnel to direct radiation from the reactor core. A storage chamber is also provided within the radiation facility to contain an irradiated sample during the period of time required to reduce the radioactivity enough to permit removal of the sample for external handling. (AEC)

CANDU in-reactor quantitative visual-based inspection techniques

NASA Astrophysics Data System (ADS)

Rochefort, P. A.

2009-02-01

This paper describes two separate visual-based inspection procedures used at CANDU nuclear power generating stations. The techniques are quantitative in nature and are delivered and operated in highly radioactive environments with access that is restrictive, and in one case is submerged. Visual-based inspections at stations are typically qualitative in nature. For example a video system will be used to search for a missing component, inspect for a broken fixture, or locate areas of excessive corrosion in a pipe. In contrast, the methods described here are used to measure characteristic component dimensions that in one case ensure ongoing safe operation of the reactor and in the other support reactor refurbishment. CANDU reactors are Pressurized Heavy Water Reactors (PHWR). The reactor vessel is a horizontal cylindrical low-pressure calandria tank approximately 6 m in diameter and length, containing heavy water as a neutron moderator. Inside the calandria, 380 horizontal fuel channels (FC) are supported at each end by integral end-shields. Each FC holds 12 fuel bundles. The heavy water primary heat transport water flows through the FC pressure tube, removing the heat from the fuel bundles and delivering it to the steam generator. The general design of the reactor governs both the type of measurements that are required and the methods to perform the measurements. The first inspection procedure is a method to remotely measure the gap between FC and other in-core horizontal components. The technique involves delivering vertically a module with a high-radiation-resistant camera and lighting into the core of a shutdown but fuelled reactor. The measurement is done using a line-of-sight technique between the components. Compensation for image perspective and viewing elevation to the measurement is required. The second inspection procedure measures flaws within the reactor's end shield FC calandria tube rolled joint area. The FC calandria tube (the outer shell of the FC) is sealed by rolling its ends into the rolled joint area. During reactor refurbishment, the original FC calandria tubes are removed, potentially scratching the rolled joint area and, thereby, compromising the seal with the new FC calandria tube. The procedure involves delivering an inspection module having a radiation-resistant camera, standard lighting, and a structured lighting projector. The surface is inspected by rotating the module within the rolled joint area. If a flaw is detected, its depth and width are gauged from the profile variation of the structured lighting in a captured image. As well, the diameter profile of the area is measured from the analysis of a series of captured circumferential images of the structured lighting profiles on the surface.

Performance and microbial community composition dynamics of aerobic granular sludge from sequencing batch bubble column reactors operated at 20 degrees C, 30 degrees C, and 35 degrees C.

PubMed

Ebrahimi, Sirous; Gabus, Sébastien; Rohrbach-Brandt, Emmanuelle; Hosseini, Maryam; Rossi, Pierre; Maillard, Julien; Holliger, Christof

2010-07-01

Two bubble column sequencing batch reactors fed with an artificial wastewater were operated at 20 degrees C, 30 degrees C, and 35 degrees C. In a first stage, stable granules were obtained at 20 degrees C, whereas fluffy structures were observed at 30 degrees C. Molecular analysis revealed high abundance of the operational taxonomic unit 208 (OTU 208) affiliating with filamentous bacteria Leptothrix spp. at 30 degrees C, an OTU much less abundant at 20 degrees C. The granular sludge obtained at 20 degrees C was used for the second stage during which one reactor was maintained at 20 degrees C and the second operated at 30 degrees C and 35 degrees C after prior gradual increase of temperature. Aerobic granular sludge with similar physical properties developed in both reactors but it had different nutrient elimination performances and microbial communities. At 20 degrees C, acetate was consumed during anaerobic feeding, and biological phosphorous removal was observed when Rhodocyclaceae-affiliating OTU 214 was present. At 30 degrees C and 35 degrees C, acetate was mainly consumed during aeration and phosphorous removal was insignificant. OTU 214 was almost absent but the Gammaproteobacteria-affiliating OTU 239 was more abundant than at 20 degrees C. Aerobic granular sludge at all temperatures contained abundantly the OTUs 224 and 289 affiliating with Sphingomonadaceae indicating that this bacterial family played an important role in maintaining stable granular structures.

Design Study of a Modular Gas-Cooled, Closed-Brayton Cycle Reactor for Marine Use

DTIC Science & Technology

1989-06-01

materials in the core and surroundings. To investigate this design point in the marine variant I developed the program HEAT.BAS to perform a one-dimensional...helium as the working fluid. The core is a graphite moderated, epithermal spectrum reactor, using TRISO fuel particles in extruded graphite fuel elements...The fuel is highly enriched U2315 . The containment is shaped in an inverted ’T’ with two sections. The upper section contains the reactor core

NUCLEAR REACTOR

DOEpatents

Young, G.

1963-01-01

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

SELF-REGULATING BOILING-WATER NUCLEAR REACTORS

DOEpatents

Ransohoff, J.A.; Plawchan, J.D.

1960-08-16

A boiling-water reactor was designed which comprises a pressure vessel containing a mass of water, a reactor core submerged within the water, a reflector tank disposed within the reactor, the reflector tank being open at the top to the interior of the pressure vessel, and a surge tank connected to the reflector tank. In operation the reflector level changes as a function of the pressure witoin the reactor so that the reactivity of the reactor is automatically controlled.

Thermal and hydraulic analysis of a cylindrical blanket module design for a tokamak reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, A.Y.

1978-10-01

Various existing blanket design concepts for a tokamak fusion reactor were evaluated and assessed. These included the demonstration power reactors of ORNL, GA and others. As a result of this study, a cylindrical, modularized blanket design concept was developed. The module is a double-walled, stainless steel 316 cylinder containing liquid lithium for tritium breeding and is cooled by pressurized helium. Steady state and transient thermal conditions under normal and some off-design conditions were analyzed and presented. At the steady state reference operating point the maximum structure temperature is 452/sup 0/C at the maximum stressed location and is 495/sup 0/C atmore » the less stressed location. The coolant inlet pressure is 54.4 atm, the inlet temperature is 200/sup 0/C and the exit temperature is 435/sup 0/C. The coolant could be utilized with a helium/steam turbine power conversion system with a cycle thermal efficiency of 30.8%.« less

78 FR 28896 - Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-16

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0095] Design Limits and Loading Combinations for Metal... Regulatory Guide (RG) 1.57, ``Design Limits and Loading Combinations for Metal Primary Reactor Containment... the NRC staff considers acceptable for design limits and loading combinations for metal primary...

MOLTEN PLUTONIUM FUELED FAST BREEDER REACTOR

DOEpatents

Kiehn, R.M.; King, L.D.P.; Peterson, R.E.; Swickard, E.O. Jr.

1962-06-26

A description is given of a nuclear fast reactor fueled with molten plutonium containing about 20 kg of plutonium in a tantalum container, cooled by circulating liquid sodium at about 600 to 650 deg C, having a large negative temperature coefficient of reactivity, and control rods and movable reflector for criticality control. (AEC)

Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Bergeron, A.; Dionne, B.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimentalmore » device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.« less

Mechanosynthesis of A Ferritic ODS (Oxide Dispersion Strengthened) Steel Containing 14% Chromium and Its Characterization

NASA Astrophysics Data System (ADS)

Rivai, A. K.; Dimyati, A.; Adi, W. A.

2017-05-01

One of the advanced materials for application at high temperatures which is aggressively developed in the world is ODS (Oxide Dispersion strengthened) steel. ODS ferritic steels are one of the candidate materials for future nuclear reactors in the world (Generation IV reactors) because it is able to be used in the reactor above 600 °C. ODS ferritic steels have also been developed for the interconnect material of SOFC (Solid Oxide Fuel Cell) which will be exposed to about 800 °C of temperature. The steel is strengthened by dispersing homogeneously of oxide particles (ceramic) in nano-meter sized in the matrix of the steel. Synthesis of a ferritic ODS steel by dispersion of nano-particles of yttrium oxide (yttria: Y2O3) as the dispersion particles, and containing high-chromium i.e. 14% has been conducted. Synthesis of the ODS steels was done mechanically (mechanosynthesis) using HEM (High Energy ball Milling) technique for 40 and 100 hours. The resulted samples were characterized using SEM-EDS (Scanning Electron Microscope-Energy Dispersive Spectroscope), and XRD (X-ray diffraction) to analyze the microstructure characteristics. The results showed that the crystal grains of the sample with 100 hours milling time was much smaller than the sample with 40 hours milling time, and some amount of alloy was formed during the milling process even for 40 hours milling time. Furthermore, the structure analysis revealed that some amount of iron atom substituted by a slight amount of chromium atom as a solid solution. The quantitative analysis showed that the phase mostly consisted of FeCr solid-solution with the structure was BCC (body-centered cubic).

Polymer-Immobilized Photosensitizers for Continuous Eradication of Bacteria

PubMed Central

Valkov, Anton; Nakonechny, Faina; Nisnevitch, Marina

2014-01-01

The photosensitizers Rose Bengal (RB) and methylene blue (MB), when immobilized in polystyrene, were found to exhibit high antibacterial activity in a continuous regime. The photosensitizers were immobilized by dissolution in chloroform, together with polystyrene, with further evaporation of the solvent, yielding thin polymeric films. Shallow reservoirs, bottom-covered with these films, were used for constructing continuous-flow photoreactors for the eradication of Gram-positive Staphylococcus aureus, Gram-negative Escherichia coli and wastewater bacteria under illumination with visible white light using a luminescent lamp at a 1.8 mW·cm−2 fluence rate. The bacterial concentration decreased by two to five orders of magnitude in separate reactors with either immobilized RB or MB, as well as in three reactors connected in series, which contained one of the photosensitizers. Bacterial eradication reached more than five orders of magnitude in two reactors connected in series, where the first reactor contained immobilized RB and the second contained immobilized MB. PMID:25158236

Two component-three dimensional catalysis

DOEpatents

Schwartz, Michael; White, James H.; Sammells, Anthony F.

2002-01-01

This invention relates to catalytic reactor membranes having a gas-impermeable membrane for transport of oxygen anions. The membrane has an oxidation surface and a reduction surface. The membrane is coated on its oxidation surface with an adherent catalyst layer and is optionally coated on its reduction surface with a catalyst that promotes reduction of an oxygen-containing species (e.g., O.sub.2, NO.sub.2, SO.sub.2, etc.) to generate oxygen anions on the membrane. The reactor has an oxidation zone and a reduction zone separated by the membrane. A component of an oxygen containing gas in the reduction zone is reduced at the membrane and a reduced species in a reactant gas in the oxidation zone of the reactor is oxidized. The reactor optionally contains a three-dimensional catalyst in the oxidation zone. The adherent catalyst layer and the three-dimensional catalyst are selected to promote a desired oxidation reaction, particularly a partial oxidation of a hydrocarbon.

Weld monitor and failure detector for nuclear reactor system

DOEpatents

Sutton, Jr., Harry G.

1987-01-01

Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

Rotating plug bearing and seal

DOEpatents

Wade, Elman E.

1977-01-01

A bearing and seal structure for nuclear reactors utilizing rotating plugs above the nuclear reactor vessel. The structure permits lubrication of bearings and seals of the rotating plugs without risk of the lubricant draining into the reactor vessel below. The structure permits lubrication by utilizing a rotating outer race bearing.

Radioactive and other environmental threats to the United States and the Arctic resulting from past Soviet activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

Earlier this year the Senate Intelligence Committee began to receive reports from environmental and nuclear scientists in Russia detailing the reckless nuclear waste disposal practices, nuclear accidents and the use of nuclear detonations. We found that information disturbing to say the least. Also troubling is the fact that 15 Chernobyl style RBMK nuclear power reactors continue to operate in the former Soviet Union today. These reactors lack a containment structure and they`re designed in such a way that nuclear reaction can actually increase when the reactor overheats. As scientists here at the University of Alaska have documented, polar air massesmore » and prevailing weather patterns provide a pathway for radioactive contaminants from Eastern Europe and Western Russia, where many of these reactors are located. The threats presented by those potential radioactive risks are just a part of a larger Arctic pollution problem. Every day, industrial activities of the former Soviet Union continue to create pollutants. I think we should face up to the reality that in a country struggling for economic survival, environment protection isn`t necessarily the high priority. And that could be very troubling news for the Arctic in the future.« less

Decontamination and deactivation of the power burst facility at the Idaho National Laboratory.

PubMed

Greene, Christy Jo

2007-05-01

Successful decontamination and deactivation of the Power Burst Facility located at the Idaho National Laboratory was accomplished through the use of extensive planning, job sequencing, engineering controls, continuous radiological support, and the use of a dedicated group of experienced workers. Activities included the removal and disposal of irradiated fuel, miscellaneous reactor components and debris stored in the canal, removal and disposition of a 15.6 curie Pu:Be start-up source, removal of an irradiated in-pile tube, and the removal of approximately 220,000 pounds of lead that was used as shielding primarily in Cubicle 13. The canal and reactor vessel were drained and water was transferred to an evaporation tank adjacent to the facility. The canal was decontaminated using underwater divers, and epoxy was affixed to the interior surfaces of the canal to contain loose contamination. The support structures and concrete or steel frame walls that form the confinement were left in place. The reactor core was left in place and a carbon steel shielding plate was placed over the reactor core to reduce radiation levels. All low-level waste and mixed low level waste generated as a result of the work activities was characterized and disposed.

Nuclear reactor with makeup water assist from residual heat removal system

DOEpatents

Corletti, Michael M.; Schulz, Terry L.

1993-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

Nuclear reactor with makeup water assist from residual heat removal system

DOEpatents

Corletti, M.M.; Schulz, T.L.

1993-12-07

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

Code of Federal Regulations, 2012 CFR

2012-01-01

... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... of the shipper, carrier, and receiver of the irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the shipment, as specified in the...

Treatment of acidic sulfate-containing wastewater using revolving algae biofilm reactors: Sulfur removal performance and microbial community characterization.

PubMed

Zhou, Haoyuan; Sheng, Yanqing; Zhao, Xuefei; Gross, Martin; Wen, Zhiyou

2018-05-18

Industries such as mining operations are facing challenges of treating sulfur-containing wastewater such as acid mine drainage (AMD) generated in their plant. The aim of this work is to evaluate the use of a revolving algal biofilm (RAB) reactor to treat AMD with low pH (3.5-4) and high sulfate content (1-4 g/L). The RAB reactors resulted in sulfate removal efficiency up to 46% and removal rate up to 0.56 g/L-day, much higher than those obtained in suspension algal culture. The high-throughput sequencing revealed that the RAB reactor contained diverse cyanobacteria, green algae, diatoms, and acid reducing bacteria that contribute the sulfate removal through various mechanisms. The RAB reactors also showed a superior performance of COD, ammonia and phosphorus removal. Collectively, the study demonstrated that RAB-based process is an effective method to remove sulfate in wastewater with small footprint and can be potentially installed in municipal or industrial wastewater treatment facilities. Copyright © 2018 Elsevier Ltd. All rights reserved.

Method and apparatus for producing synthesis gas

DOEpatents

Hemmings, John William; Bonnell, Leo; Robinson, Earl T.

2010-03-03

A method and apparatus for reacting a hydrocarbon containing feed stream by steam methane reforming reactions to form a synthesis gas. The hydrocarbon containing feed is reacted within a reactor having stages in which the final stage from which a synthesis gas is discharged incorporates expensive high temperature materials such as oxide dispersed strengthened metals while upstream stages operate at a lower temperature allowing the use of more conventional high temperature alloys. Each of the reactor stages incorporate reactor elements having one or more separation zones to separate oxygen from an oxygen containing feed to support combustion of a fuel within adjacent combustion zones, thereby to generate heat to support the endothermic steam methane reforming reactions.

Passive containment cooling system with drywell pressure regulation for boiling water reactor

DOEpatents

Hill, Paul R.

1994-01-01

A boiling water reactor having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit.

Control of reactor coolant flow path during reactor decay heat removal

DOEpatents

Hunsbedt, Anstein N.

1988-01-01

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.

Progressing batch hydrolysis process

DOEpatents

Wright, J.D.

1985-01-10

A progressive batch hydrolysis process is disclosed for producing sugar from a lignocellulosic feedstock. It comprises passing a stream of dilute acid serially through a plurality of percolation hydrolysis reactors charged with feed stock, at a flow rate, temperature and pressure sufficient to substantially convert all the cellulose component of the feed stock to glucose. The cooled dilute acid stream containing glucose, after exiting the last percolation hydrolysis reactor, serially fed through a plurality of pre-hydrolysis percolation reactors, charged with said feedstock, at a flow rate, temperature and pressure sufficient to substantially convert all the hemicellulose component of said feedstock to glucose. The dilute acid stream containing glucose is cooled after it exits the last prehydrolysis reactor.

Progressing batch hydrolysis process

DOEpatents

Wright, John D.

1986-01-01

A progressive batch hydrolysis process for producing sugar from a lignocellulosic feedstock, comprising passing a stream of dilute acid serially through a plurality of percolation hydrolysis reactors charged with said feedstock, at a flow rate, temperature and pressure sufficient to substantially convert all the cellulose component of the feedstock to glucose; cooling said dilute acid stream containing glucose, after exiting the last percolation hydrolysis reactor, then feeding said dilute acid stream serially through a plurality of prehydrolysis percolation reactors, charged with said feedstock, at a flow rate, temperature and pressure sufficient to substantially convert all the hemicellulose component of said feedstock to glucose; and cooling the dilute acid stream containing glucose after it exits the last prehydrolysis reactor.

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael A. Pope

2011-10-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physicsmore » design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.« less

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francesco Venneri; Chang-Keun Jo; Jae-Man Noh

2010-09-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physicsmore » design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.« less

Improved Recovery and Identification of Membrane Proteins from Rat Hepatic Cells using a Centrifugal Proteomic Reactor*

PubMed Central

Zhou, Hu; Wang, Fangjun; Wang, Yuwei; Ning, Zhibin; Hou, Weimin; Wright, Theodore G.; Sundaram, Meenakshi; Zhong, Shumei; Yao, Zemin; Figeys, Daniel

2011-01-01

Despite their importance in many biological processes, membrane proteins are underrepresented in proteomic analysis because of their poor solubility (hydrophobicity) and often low abundance. We describe a novel approach for the identification of plasma membrane proteins and intracellular microsomal proteins that combines membrane fractionation, a centrifugal proteomic reactor for streamlined protein extraction, protein digestion and fractionation by centrifugation, and high performance liquid chromatography-electrospray ionization-tandem MS. The performance of this approach was illustrated for the study of the proteome of ER and Golgi microsomal membranes in rat hepatic cells. The centrifugal proteomic reactor identified 945 plasma membrane proteins and 955 microsomal membrane proteins, of which 63 and 47% were predicted as bona fide membrane proteins, respectively. Among these proteins, >800 proteins were undetectable by the conventional in-gel digestion approach. The majority of the membrane proteins only identified by the centrifugal proteomic reactor were proteins with ≥2 transmembrane segments or proteins with high molecular mass (e.g. >150 kDa) and hydrophobicity. The improved proteomic reactor allowed the detection of a group of endocytic and/or signaling receptor proteins on the plasma membrane, as well as apolipoproteins and glycerolipid synthesis enzymes that play a role in the assembly and secretion of apolipoprotein B100-containing very low density lipoproteins. Thus, the centrifugal proteomic reactor offers a new analytical tool for structure and function studies of membrane proteins involved in lipid and lipoprotein metabolism. PMID:21749988

Application of biocatalysts to Space Station ECLSS and PMMS water reclamation

NASA Technical Reports Server (NTRS)

Jolly, Clifford D.; Bagdigian, Robert M.

1989-01-01

Immobilized enzyme reactors have been developed and tested for potential water reclamation applications in the Space Station Freedom Environmental Control and Life Support System (ECLSS) and Process Materials Management System (PMMS). The reactors convert low molecular weight organic contaminants found in ECLSS and PMMS wastewaters to compounds that are more efficiently removed by existing technologies. Demonstration of the technology was successfully achieved with two model reactors. A packed bed reactor containing immobilized urease was found to catalyze the complete decomposition of urea to by-products that were subsequently removed using conventional ion exchange results. A second reactor containing immobilized alcohol oxidase showed promising results relative to its ability to convert methanol and ethanol to the corresponding aldehydes for subsequent removal. Preliminary assessments of the application of biocatalysts to ECLSS and PMMS water reclamation sytems are presented.

AHTR Mechanical, Structural, and Neutronic Preconceptual Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Varma, V.K.; Holcomb, D.E.; Peretz, F.J.

2012-09-15

This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming an option for commercial reactor deployment. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The reactor concept development remains at a preconceptual levelmore » of maturity. While the overall appearance of an AHTR design is anticipated to be similar to the current concept, optimized dimensions will differ from those presented here. The AHTR employs plate type coated particle fuel assemblies with rapid, off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month two-batch cycle with 9 wt. % enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The report includes a preconceptual design of the manipulators, the fuel transfer system, and the used fuel storage system. The present design intent is for used fuel to be stored inside of containment for at least six months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design concept incorporates fully passive responses to all identified design basis or non-very-low frequency beyond design basis accidents as well as multiple levels of radioactive material containment. Key building design elements include (1) below grade siting to minimize vulnerability to aircraft impact, (2) multiple natural circulation decay heat rejection chimneys, (3) seismic base isolation, and (4) decay heat powered back-up electricity generation.« less

AHTR Mechanical, Structural, And Neutronic Preconceptual Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Varma, Venugopal Koikal; Holcomb, David Eugene; Peretz, Fred J

2012-10-01

This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming a commercial reactor class. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The AHTR employs plate type coated particle fuel assemblies with rapid,more » off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month 2-batch cycle with 9 weight-percent enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The present design intent is for used fuel to be stored inside of containment for at least 6 months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design concept incorporates multiple levels of radioactive material containment including fully passive responses to all identified design basis or non-very-low frequency beyond design basis accidents. Key building design elements include: 1) below grade siting to minimize vulnerability to aircraft impact, 2) multiple natural circulation decay heat rejection chimneys, 3) seismic base isolation, and 4) decay heat powered back-up electricity generation. The report provides a preconceptual design of the manipulators, the fuel transfer system, and the salt transfer loops. The mechanical handling of the fuel and how it is accomplished without instrumentation inside the salt is described within the report. All drives for the manipulators reside outside the reactor top flange. The design has also taken into account the transportability of major components and how they will be assembled on site« less

Passive containment cooling system

DOEpatents

Conway, Lawrence E.; Stewart, William A.

1991-01-01

A containment cooling system utilizes a naturally induced air flow and a gravity flow of water over the containment shell which encloses a reactor core to cool reactor core decay heat in two stages. When core decay heat is greatest, the water and air flow combine to provide adequate evaporative cooling as heat from within the containment is transferred to the water flowing over the same. The water is heated by heat transfer and then evaporated and removed by the air flow. After an initial period of about three to four days when core decay heat is greatest, air flow alone is sufficient to cool the containment.

Molecular structures and thermodynamic properties of monohydrated gaseous iodine compounds: Modelling for severe accident simulation

NASA Astrophysics Data System (ADS)

Sudolská, Mária; Cantrel, Laurent; Budzák, Šimon; Černušák, Ivan

2014-03-01

Monohydrated complexes of iodine species (I, I2, HI, and HOI) have been studied by correlated ab initio calculations. The standard enthalpies of formation, Gibbs free energy and the temperature dependence of the heat capacities at constant pressure were calculated. The values obtained have been implemented in ASTEC nuclear accident simulation software to check the thermodynamic stability of hydrated iodine compounds in the reactor coolant system and in the nuclear containment building of a pressurised water reactor during a severe accident. It can be concluded that iodine complexes are thermodynamically unstable by means of positive Gibbs free energies and would be represented by trace level concentrations in severe accident conditions; thus it is well justified to only consider pure iodine species and not hydrated forms.

Acoustic emission signal processing technique to characterize reactor in-pile phenomena

DOE Office of Scientific and Technical Information (OSTI.GOV)

Agarwal, Vivek, E-mail: vivek.agarwal@inl.gov; Tawfik, Magdy S., E-mail: magdy.tawfik@inl.gov; Smith, James A., E-mail: james.smith@inl.gov

2015-03-31

Existing and developing advanced sensor technologies and instrumentation will allow non-intrusive in-pile measurement of temperature, extension, and fission gases when coupled with advanced signal processing algorithms. The transmitted measured sensor signals from inside to the outside of containment structure are corrupted by noise and are attenuated, thereby reducing the signal strength and the signal-to-noise ratio. Identification and extraction of actual signal (representative of an in-pile phenomenon) is a challenging and complicated process. In the paper, empirical mode decomposition technique is utilized to reconstruct actual sensor signal by partially combining intrinsic mode functions. Reconstructed signal will correspond to phenomena and/or failuremore » modes occurring inside the reactor. In addition, it allows accurate non-intrusive monitoring and trending of in-pile phenomena.« less

Measurement and Analysis of Structural Integrity of Reactor Core Support Structure in Pressurized Water Reactor (PWR) Plant

NASA Astrophysics Data System (ADS)

Ansari, Saleem A.; Haroon, Muhammad; Rashid, Atif; Kazmi, Zafar

2017-02-01

Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS.

Combined on-board hydride slurry storage and reactor system and process for hydrogen-powered vehicles and devices

DOEpatents

Brooks, Kriston P; Holladay, Jamelyn D; Simmons, Kevin L; Herling, Darrell R

2014-11-18

An on-board hydride storage system and process are described. The system includes a slurry storage system that includes a slurry reactor and a variable concentration slurry. In one preferred configuration, the storage system stores a slurry containing a hydride storage material in a carrier fluid at a first concentration of hydride solids. The slurry reactor receives the slurry containing a second concentration of the hydride storage material and releases hydrogen as a fuel to hydrogen-power devices and vehicles.

Production of polyhydroxyalkanoates (PHA) by bacterial consortium from excess sludge fermentation liquid at laboratory and pilot scales.

PubMed

Jia, Qianqian; Xiong, Huilei; Wang, Hui; Shi, Hanchang; Sheng, Xinying; Sun, Run; Chen, Guoqiang

2014-11-01

The generation of polyhydroxyalkanoates (PHA) from excess sludge fermentation liquid (SFL) was studied at lab and pilot scale. A PHA-accumulated bacterial consortium (S-150) was isolated from activated sludge using simulated SFL (S-SFL) contained high concentration volatile fatty acids (VFA) and nitrogen. The maximal PHA content accounted for 59.18% in S-SFL and dropped to 23.47% in actual SFL (L-SFL) of the dry cell weight (DCW) at lab scale. The pilot-scale integrated system comprised an anaerobic fermentation reactor (AFR), a ceramic membrane system (CMS) and a PHA production bio-reactor (PHAR). The PHA content from pilot-scale SFL (P-SFL) finally reached to 59.47% DCW with the maximal PHA yield coefficient (YP/S) of 0.17 g PHA/g COD. The results indicated that VFA-containing SFL was suitable for PHA production. The adverse impact of excess nitrogen and non-VFAs in SFL might be eliminated by pilot-scale domestication, which might resulted in community structure optimization and substrate selective ability improvement of S-150. Copyright © 2014 Elsevier Ltd. All rights reserved.

Tower Shielding Reactor II design and operation report: Vol. 2. Safety Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holland, L. B.; Kolb, J. O.

1970-01-01

Information on the Tower Shielding Reactor II is contained in the TSR-II Design and Operation Report and in the Tower Shielding Facility Manual. The TSR-II Design and Operating Report consists of three volumes. Volume 1 is Descriptions of the Tower Shielding Reactor II and Facility; Volume 2 is Safety analysis of the Tower Shielding Reactor II; and Volume 3 is the Assembly and Testing of the Tower Shielding Reactor II Control Mechanism Housing.

PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP

DOEpatents

Puechl, K.H.

1963-09-24

A pressurized water reactor is described having a core containing Pu/sup 240/ in which the effective microscopic neutronabsorption cross section of Pu/sup 240/ in unconverted condition decreases as the time of operation of the reactor increases, in order to compensate for loss of reactivity resulting from fission product buildup during reactor operation. This means serves to improve the efficiency of the reactor operation by reducing power losses resulting from control rods and burnable poisons. (AEC)

Digital computer operation of a nuclear reactor

DOEpatents

Colley, R.W.

1982-06-29

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.

Digital computer operation of a nuclear reactor

DOEpatents

Colley, Robert W.

1984-01-01

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.

Passive containment cooling system with drywell pressure regulation for boiling water reactor

DOEpatents

Hill, P.R.

1994-12-27

A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

NON-CORROSIVE PLUTONIUM FUEL SYSTEMS

DOEpatents

Coffinberry, A.S.; Waber, J.T.

1962-10-23

An improved plutonium reactor liquid fuel is described for utilization in a nuclear reactor having a tantalum fuel containment vessel. The fuel consists of plutonium and a diluent such as iron, cobalt, nickel, cerium, cerium-- iron, cerium--cobalt, cerium--nickel, and cerium--copper, and an additive of carbon and silicon. The carbon and silicon react with the tantalum container surface to form a coating that is self-healing and prevents the corrosive action of liquid plutonium on the said tantalum container. (AEC)

Synthesis gas method and apparatus

DOEpatents

Kelly, Sean M.; Kromer, Brian R.; Litwin, Michael M.; Rosen, Lee J.; Christie, Gervase Maxwell; Wilson, Jamie; Kosowski, Lawrence W; Robinson, Charles

2015-11-06

A method and apparatus for producing a synthesis gas product having one or more oxygen transport membrane elements thermally coupled to one or more catalytic reactors such that heat generated from the oxygen transport membrane element supplies endothermic heating requirements for steam methane reforming reactions occurring within the catalytic reactor through radiation and convention heat transfer. A hydrogen containing stream containing no more than 20 percent methane is combusted within the oxygen transport membrane element to produce the heat and a heated combustion product stream. The heated combustion product stream is combined with a reactant stream to form a combined stream that is subjected to the reforming within the catalytic reactor. The apparatus may include modules in which tubular membrane elements surround a central reactor tube.

Synthesis gas method and apparatus

DOEpatents

Kelly, Sean M.; Kromer, Brian R.; Litwin, Michael M.; Rosen, Lee J.; Christie, Gervase Maxwell; Wilson, Jamie R.; Kosowski, Lawrence W.; Robinson, Charles

2013-01-08

A method and apparatus for producing a synthesis gas product having one or more oxygen transport membrane elements thermally coupled to one or more catalytic reactors such that heat generated from the oxygen transport membrane element supplies endothermic heating requirements for steam methane reforming reactions occurring within the catalytic reactor through radiation and convention heat transfer. A hydrogen containing stream containing no more than 20 percent methane is combusted within the oxygen transport membrane element to produce the heat and a heated combustion product stream. The heated combustion product stream is combined with a reactant stream to form a combined stream that is subjected to the reforming within the catalytic reactor. The apparatus may include modules in which tubular membrane elements surround a central reactor tube.

BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hetrick, D.L.; Sowers, G.W.

1978-06-01

This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. Amore » list of variable names and a listing for BRENDA are included as appendices.« less

REACTOR SHIELD

DOEpatents

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

1959-02-17

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

System and process for the production of syngas and fuel gasses

DOEpatents

Bingham, Dennis N.; Kllingler, Kerry M.; Turner, Terry D.; Wilding, Bruce M.; Benefiel, Bradley C.

2014-04-01

The production of gasses and, more particularly, to systems and methods for the production of syngas and fuel gasses including the production of hydrogen are set forth. In one embodiment system and method includes a reactor having a molten pool of a material comprising sodium carbonate. A supply of conditioned water is in communication with the reactor. A supply of carbon containing material is also in communication with the reactor. In one particular embodiment, the carbon containing material may include vacuum residuum (VR). The water and VR may be kept at desired temperatures and pressures compatible with the process that is to take place in the reactor. When introduced into the reactor, the water, the VR and the molten pool may be homogenously mixed in an environment in which chemical reactions take place including the production of hydrogen and other gasses.

System and process for the production of syngas and fuel gasses

DOEpatents

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

2015-04-21

The production of gasses and, more particularly, to systems and methods for the production of syngas and fuel gasses including the production of hydrogen are set forth. In one embodiment system and method includes a reactor having a molten pool of a material comprising sodium carbonate. A supply of conditioned water is in communication with the reactor. A supply of carbon containing material is also in communication with the reactor. In one particular embodiment, the carbon containing material may include vacuum residuum (VR). The water and VR may be kept at desired temperatures and pressures compatible with the process that is to take place in the reactor. When introduced into the reactor, the water, the VR and the molten pool may be homogenously mixed in an environment in which chemical reactions take place including the production of hydrogen and other gasses.

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baratta, A.J.

1997-07-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts andmore » engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.« less

NEUTRONIC REACTOR SYSTEM

DOEpatents

Goett, J.J.

1961-01-24

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

Status report on the cold neutron source of the Garching neutron research facility FRM-II

NASA Astrophysics Data System (ADS)

Gobrecht, K.; Gutsmiedl, E.; Scheuer, A.

2002-01-01

The new high flux research reactor of the Technical University of Munich (Technische Universität München, TUM) will be equipped with a cold neutron source (CNS). The centre of the CNS will be located in the D 2O-reflector tank at 400 mm from the reactor core axis close to the thermal neutron flux maximum. The power of 4500 W developed by the nuclear heating in the 16 l of liquid deuterium at 25 K, and in the structures, is evacuated by a two-phase thermal siphon avoiding film boiling and flooding. The thermal siphon is a single tube with counter current flow. It is inclined by 10° from vertical, and optimised for a deuterium flow rate of 14 g/s. Optimisation of structure design and material, as well as safety aspects will be discussed. Those parts of the structure, which are exposed to high thermal neutron flux, are made from Zircaloy 4 and 6061T6 aluminium. Structure failure due to embrittlement of the structure material under high rapid neutron flux is very improbable during the lifetime of the CNS (30 years). Double, in pile even triple, containment with inert gas liner guarantees lack of explosion risk and of tritium contamination to the environment. Adding a few percent of hydrogen (H 2) to the deuterium (D 2) will improve the moderating properties of our relatively small moderator volume. Nearly all of the hydrogen is bound in the form of HD molecules. A long-term change of the hydrogen content in the deuterium is avoided by storing the mixture not in a gas buffer volume but as a metal hydride at low pressure. The metal hydride storage system contains two getter beds, one with 250 kg of LaCo 3Ni 2, the other one with 150 kg of ZrCo 0.8Ni 0.2. Each bed can take the total gas inventory, both beds together can absorb the total gas inventory in <6 min at a pressure <3 bar. The new reactor will have 13 beam tubes, 4 of which are looking at the CNS, including two for very cold (VCN) and ultra-cold neutron (UCN) production. The latter will take place in the horizontal beam tube SR4, which will house an additional cryogenic moderator (e.g. solid deuterium). More than 60% of the experiments foreseen in the new neutron research facility will use cold neutrons from the CNS. The mounting of the hardware components of the CNS into the reactor has started in the spring of 2000. The CNS went into trial operation in the end of year 2000.

Nuclear reactor pressure vessel support system

DOEpatents

Sepelak, George R.

1978-01-01

A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

Synthesis of carbohydrates in a continuous flow reactor by immobilized phosphatase and aldolase.

PubMed

Babich, Lara; Hartog, Aloysius F; van Hemert, Lieke J C; Rutjes, Floris P J T; Wever, Ron

2012-12-01

Herein, we report a new flow process with immobilized enzymes to synthesize complex chiral carbohydrate analogues from achiral inexpensive building blocks in a three-step cascade reaction. The first reactor contained immobilized acid phosphatase, which phosphorylated dihydroxyacetone to dihydroxyacetone phosphate using pyrophosphate as the phosphate donor. The second flow reactor contained fructose-1,6-diphosphate aldolase (RAMA, rabbit muscle aldolase) or rhamnulose-1-phosphate aldolase (RhuA from Thermotoga maritima) and acid phosphatase. The immobilized aldolases coupled the formed dihydroxyacetone phosphate to aldehydes, resulting in phosphorylated carbohydrates. A final reactor containing acid phosphatase that dephosphorylated the phosphorylated product yielded the final product. Different aldehydes were used to synthesize carbohydrates on a gram scale. To demonstrate the feasibility of the flow systems, we synthesized 0.6 g of the D-fagomine precursor. By using immobilized aldolase RhuA we were also able to obtain other stereoisomers of the D-fagomine precursor. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Oxygen transport membrane reactor based method and system for generating electric power

DOEpatents

Kelly, Sean M.; Chakravarti, Shrikar; Li, Juan

2017-02-07

A carbon capture enabled system and method for generating electric power and/or fuel from methane containing sources using oxygen transport membranes by first converting the methane containing feed gas into a high pressure synthesis gas. Then, in one configuration the synthesis gas is combusted in oxy-combustion mode in oxygen transport membranes based boiler reactor operating at a pressure at least twice that of ambient pressure and the heat generated heats steam in thermally coupled steam generation tubes within the boiler reactor; the steam is expanded in steam turbine to generate power; and the carbon dioxide rich effluent leaving the boiler reactor is processed to isolate carbon. In another configuration the synthesis gas is further treated in a gas conditioning system configured for carbon capture in a pre-combustion mode using water gas shift reactors and acid gas removal units to produce hydrogen or hydrogen-rich fuel gas that fuels an integrated gas turbine and steam turbine system to generate power. The disclosed method and system can also be adapted to integrate with coal gasification systems to produce power from both coal and methane containing sources with greater than 90% carbon isolation.

Organic matter and containment of uranium and fissiogenic isotopes at the Oklo natural reactors

USGS Publications Warehouse

Nagy, B.; Gauthier-Lafaye, F.; Holliger, P.; Davis, D.W.; Mossman, D.J.; Leventhal, J.S.; Rigali, M.J.; Parnell, J.

1991-01-01

SOME of the Precambrian natural fission reactors at Oklo in Gabon contain abundant organic matter1,2, part of which was liquefied at the time of criticality and subsequently converted to a graphitic solid3,4. The liquid organic matter helps to reduce U(VI) to U(IV) from aqueous solutions, resulting in the precipitation of uraninite5. It is known that in the prevailing reactor environments, precipitated uraninite grains incorporated fission products. We report here observations which show that these uraninite crystals were held immobile within the resolidified, graphitic bitumen. Unlike water-soluble (humic) organic matter, the graphitic bituminous organics at Oklo thus enhanced radionu-clide containment. Uraninite encased in solid graphitic matter in the organic-rich reactor zones lost virtually no fissiogenic lan-thanide isotopes. The first major episode of uranium and lead migration was caused by the intrusion of a swarm of adjacent dolerite dykes about 1,100 Myr after the reactors went critical. Our results from Oklo imply that the use of organic, hydrophobic solids such as graphitic bitumen as a means of immobilizing radionuclides in pretreated nuclear waste warrants further investigation. ?? 1991 Nature Publishing Group.

Structural materials issues for the next generation fission reactors

NASA Astrophysics Data System (ADS)

Chant, I.; Murty, K. L.

2010-09-01

Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1982-09-01

This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report.

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with...

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with...

Biological hydrogen production by Clostridium acetobutylicum in an unsaturated flow reactor.

PubMed

Zhang, Husen; Bruns, Mary Ann; Logan, Bruce E

2006-02-01

A mesophilic unsaturated flow (trickle bed) reactor was designed and tested for H2 production via fermentation of glucose. The reactor consisted of a column packed with glass beads and inoculated with a pure culture (Clostridium acetobutylicum ATCC 824). A defined medium containing glucose was fed at a flow rate of 1.6 mL/min (0.096 L/h) into the capped reactor, producing a hydraulic retention time of 2.1 min. Gas-phase H2 concentrations were constant, averaging 74 +/- 3% for all conditions tested. H2 production rates increased from 89 to 220 mL/hL of reactor when influent glucose concentrations were varied from 1.0 to 10.5 g/L. Specific H2 production rate ranged from 680 to 1270 mL/g glucose per liter of reactor (total volume). The H2 yield was 15-27%, based on a theoretical limit by fermentation of 4 moles of H2 from 1 mole of glucose. The major fermentation by-products in the liquid effluent were acetate and butyrate. The reactor rapidly (within 60-72 h) became clogged with biomass, requiring manual cleaning of the system. In order to make long-term operation of the reactor feasible, biofilm accumulation in the reactor will need to be controlled through some process such as backwashing. These tests using an unsaturated flow reactor demonstrate the feasibility of the process to produce high H2 gas concentrations in a trickle-bed type of reactor. A likely application of this reactor technology could be H2 gas recovery from pre-treatment of high carbohydrate-containing wastewaters.

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoon, Dhongik S.; Jo, HangJin; Fu, Wen

A multi-application small light water reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s abilitymore » to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with CFD simulation for confirmation. Furthermore, the corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.« less

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

DOE PAGES

Yoon, Dhongik S.; Jo, HangJin; Fu, Wen; ...

2017-05-23

A multi-application small light water reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s abilitymore » to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with CFD simulation for confirmation. Furthermore, the corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.« less

Needs of Accurate Prompt and Delayed γ-spectrum and Multiplicity for Nuclear Reactor Designs

NASA Astrophysics Data System (ADS)

Rimpault, G.; Bernard, D.; Blanchet, D.; Vaglio-Gaudard, C.; Ravaux, S.; Santamarina, A.

The local energy photon deposit must be accounted accurately for Gen-IV fast reactors, advanced light-water nuclear reactors (Gen-III+) and the new experimental Jules Horowitz Reactor (JHR). The γ energy accounts for about 10% of the total energy released in the core of a thermal or fast reactor. The γ-energy release is much greater in the core of the reactor than in its structural sub-assemblies (such as reflector, control rod followers, dummy sub-assemblies). However, because of the propagation of γ from the core regions to the neighboring fuel-free assemblies, the contribution of γ energy to the total heating can be dominant. For reasons related to their performance, power reactors require a 7.5% (1σ) uncertainty for the energy deposition in non-fuelled zones. For the JHR material-testing reactor, a 5% (1 s) uncertainty is required in experimental positions. In order to verify the adequacy of the calculation of γ-heating, TLD and γ-fission chambers were used to derive the experimental heating values. Experimental programs were and are still conducted in different Cadarache facilities such as MASURCA (for SFR), MINERVE and EOLE (for JHR and Gen-III+ reactors). The comparison of calculated and measured γ-heating values shows an underestimation in all experimental programs indicating that for the most γ-production data from 239Pu in current nuclear-data libraries is highly suspicious.The first evaluation priority is for prompt γ-multiplicity for U and Pu fission but similar values for otheractinides such as Pu and U are also required. The nuclear data library JEFF3.1.1 contains most of the photon production data. However, there are some nuclei for which there are missing or erroneous data which need to be completed or modified. A review of the data available shows a lack of measurements for conducting serious evaluation efforts. New measurements are needed to guide new evaluation efforts which benefit from consolidated modeling techniques.

Reactor water cleanup system

DOEpatents

Gluntz, Douglas M.; Taft, William E.

1994-01-01

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

Sodium leak detection system for liquid metal cooled nuclear reactors

DOEpatents

Modarres, Dariush

1991-01-01

A light source is projected across the gap between the containment vessel and the reactor vessel. The reflected light is then analyzed with an absorption spectrometer. The presence of any sodium vapor along the optical path results in a change of the optical transmissivity of the media. Since the absorption spectrum of sodium is well known, the light source is chosen such that the sensor is responsive only to the presence of sodium molecules. The optical sensor is designed to be small and require a minimum of amount of change to the reactor containment vessel.

FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

DOEpatents

Abbott, W.E.; Balent, R.

1958-09-16

A fuel element design to facilitate breeding reactor fuel is described. The fuel element is comprised of a coatainer, a central core of fertile material in the container, a first bonding material surrounding the core, a sheet of fissionable material immediately surrounding the first bonding material, and a second bonding material surrounding the fissionable material and being in coniact with said container.

Recovery of elemental sulphur from anaerobic effluents through the biological oxidation of sulphides.

PubMed

de Sousa, José Tavares; Lima, Jéssyca de Freitas; da Silva, Valquíria Cordeiro; Leite, Valderi Duarte; Lopes, Wilton Silva

2017-03-01

The aim of the present study was to evaluate the biological oxidation of sulphide in two different UASB reactors by assessing the occurrence of oxidized forms of sulphur in the effluents and the amount of S 0 that could be recovered in the process. The bioreactors employed were an anaerobic hybrid (AH) reactor employing porous polyurethane foam as support media and a micro-aerated UASB reactor equipped with an aeration device above the digestion zone. The AH reactor produced a final effluent containing low concentrations of S 2- (3.87% of total sulphur load). It was achieved due to a complete oxidation of 56.1% of total sulphur. The partial biological oxidation that occurred in the AH reactor allowed the recovery of 30% of the sulphur load as S 0 . The effluent from the micro-aerated UASB reactor contained 5% of the sulphur load in the form of S 2- , while 20.9% was present as dissolved SO 4 2- and 46% was precipitated as S 0 . It is concluded that the AH reactor or micro-aeration carried out above the digestion zone of the UASB reactor favoured the biological oxidation of S 2- and the release of odourless effluents. Both technologies represent feasible and low-cost alternatives for the anaerobic treatment of domestic sewage.

Aging management program of the reactor building concrete at Point Lepreau Generating Station

NASA Astrophysics Data System (ADS)

Aldea, C.-M.; Shenton, B.; Demerchant, M. M.; Gendron, T.

2011-04-01

In order for New Brunswick Power Nuclear (NBPN) to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS) the development of an aging management plan (AMP) was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

Joule-Heated Molten Regolith Electrolysis Reactor Concepts for Oxygen and Metals Production on the Moon and Mars

NASA Technical Reports Server (NTRS)

Sibille, Laurent; Dominguez, Jesus A.

2012-01-01

The technology of direct electrolysis of molten lunar regolith to produce oxygen and molten metal alloys has progressed greatly in the last few years. The development of long-lasting inert anodes and cathode designs as well as techniques for the removal of molten products from the reactor has been demonstrated. The containment of chemically aggressive oxide and metal melts is very difficult at the operating temperatures ca. 1600 C. Containing the molten oxides in a regolith shell can solve this technical issue and can be achieved by designing a Joule-heated (sometimes called 'self-heating') reactor in which the electrolytic currents generate enough Joule heat to create a molten bath. Solutions obtained by multiphysics modeling allow the identification of the critical dimensions of concept reactors.

A peculiar segmented flow microfluidics for isoquercitrin biosynthesis based on coupling of reaction and separation.

PubMed

Gong, An; Gu, Shuang-Shuang; Wang, Jun; Sheng, Sheng; Wu, Fu-An

2015-10-01

A segmented flow containing a buffer-ionic liquid/solvent in a micro-channel reactor was applied to synthesize isoquercitrin by the hesperidinase-catalyzed selective hydrolysis of rutin, based on a novel system of reaction coupling with separation. Within the developed microchannel reactor with one T-shaped inlet and outlet, the maximum isoquercitrin yield (101.7 ± 2.6%) was achieved in 20 min at 30 °C and 4 μL/min. Compared with a continuous-flow reactor, reaction rate was increased 4-fold due to a glycine-sodium hydroxide:[Bmim][BF4]/glycerol triacetate (1:1, v/v) system that formed a slug flow in microchannel and significantly increased mass transfer rates. The mass transfer coefficient significantly increased and exhibited a linear relationship with the flow rate. Hesperidinase could be efficiently reused at least 5 times, without losing any activity. The bonding mechanism and secondary structure of hesperidinase indicated that hesperidinase had a greater affinity to rutin at a production rate of 4 μL/min in this segmented flow microreactor. Copyright © 2015 Elsevier Ltd. All rights reserved.

Analysis of trichloroethylene removal and bacterial community function based on pH-adjusted in an upflow anaerobic sludge blanket reactor.

PubMed

Zhang, Ying; Hu, Miao; Li, Pengfei; Wang, Xin; Meng, Qingjuan

2015-11-01

The study reported the upflow anaerobic sludge blanket (UASB) reactor performance in treating wastewater containing trichloroethylene (TCE) and characterized variations of bacteria composition and structure by changing the pH from 6.0 to 8.0. A slightly acidic environment (pH < 7.0) had a greater impact on the TCE removal. Illumina pyrosequencing was applied to investigate the bacterial community changes in response to pH shifts. The results demonstrated that pH greatly influenced the dominance and presence of specific populations. The potential TCE degradation pathway in the UASB reactor was proposed. Importantly, the genus Dehalobacter which was capable of reductively dechlorinating TCE was detected, and it was not found at pH of 6.0, which presumably is the reason why the removal efficiency of TCE was the lowest (80.73 %). Through Pearson correlation analyses, the relative abundance of Dehalobacter positively correlated with TCE removal efficiency (R = 0.912). However, the relative abundance of Lactococcus negatively correlated with TCE removal efficiency according to the results from Pearson correlation analyses and redundancy analysis (RDA).

Analysis of dpa rates in the HFIR reactor vessel using a hybrid Monte Carlo/deterministic method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blakeman, Edward

2016-01-01

The Oak Ridge High Flux Isotope Reactor (HFIR), which began full-power operation in 1966, provides one of the highest steady-state neutron flux levels of any research reactor in the world. An ongoing vessel integrity analysis program to assess radiation-induced embrittlement of the HFIR reactor vessel requires the calculation of neutron and gamma displacements per atom (dpa), particularly at locations near the beam tube nozzles, where radiation streaming effects are most pronounced. In this study we apply the Forward-Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS) technique in the ADVANTG code to develop variance reduction parameters for use in the MCNP radiationmore » transport code. We initially evaluated dpa rates for dosimetry capsule locations, regions in the vicinity of the HB-2 beamline, and the vessel beltline region. We then extended the study to provide dpa rate maps using three-dimensional cylindrical mesh tallies that extend from approximately 12 below to approximately 12 above the axial extent of the core. The mesh tally structures contain over 15,000 mesh cells, providing a detailed spatial map of neutron and photon dpa rates at all locations of interest. Relative errors in the mesh tally cells are typically less than 1%.« less

Evaluation of Municipal Wastewater Treatment Plant Activated Sludge for Biodegradation of Propylene Glycol as an Aircraft Deicing Fluid

DTIC Science & Technology

2012-03-01

Propylene Glycol Deicer Biodegredation Kinetics: Complete-Mix Stirred Tank Reactors , Filter, and Fluidized Bed . Journal of Environmental...scale sequencing batch reactor containing municipal waste water treatment facility activated sludge (AS) performing simultaneous organic carbon...Sequencing Batch Reactor Operation ..................................................................... 13 PG extraction from AS

ETR BUILDING, TRA642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. SOUTH SIDE ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR BUILDING, TRA-642, INTERIOR. CONSOLE FLOOR, SOUTH HALF. SOUTH SIDE OF ETR REACTOR, CAMERA FACING NORTH. CABINET CONTAINING "NUCLEAR INSTRUMENT SYSTEMS" IS RESTRICTED. INL NEGATIVE NO. HD46-18-4. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

REACTOR PHYSICS QUARTERLY REPORT JANUARY, FEBRUARY, MARCH 1970

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schmid, L. C.; Clayton, E. D.; Heineman, R. E.

1970-05-01

The objective of the Reactor Physics Quarterly Report is to inform the scientific community in a timely manner of the technical progress made on the many phases of reactor physics work within the laboratory. The report contains brief technical discussions of accomplishments in all areas where significant progress has been made during the quarter.

Archaeal and Bacterial Community Structure in an Anaerobic Digestion Reactor (Lagoon Type) Used for Biogas Production at a Pig Farm.

PubMed

Pampillón-González, Liliana; Ortiz-Cornejo, Nadia L; Luna-Guido, Marco; Dendooven, Luc; Navarro-Noya, Yendi E

2017-01-01

Biogas production from animal waste is an economically viable way to reduce environmental pollution and produce valuable products, i.e., methane and a nutrient-rich organic waste product. An anaerobic digestion reactor for biogas production from pig waste was sampled at the entrance, middle (digestion chamber), and exit of a digester, while the bacterial and archaeal community structure was studied by 16S rRNA gene metagenomics. The number of bacterial operational taxonomic units (OTU)-97% was 3-7 times larger than that of archaeal ones. Bacteria and Archaea found in feces of animals (e.g., Clostridiaceae, Lachnospiraceae, Ruminococcaceae, Methanosarcina, Methanolobus, Methanosaeta, and Methanospirillum) dominated the entrance of the digester. The digestion chamber was dominated by anaerobic sugar-fermenting OP9 bacteria and the syntrophic bacteria Candidatus Cloacamonas (Waste Water of Evry 1; WWE1). The methanogens dominant in the digestion chamber were the acetoclastic Methanosaeta and the hydrogenothrophic Methanoculleus and Methanospirillum. Similar bacterial and archaeal groups that dominated in the middle of the digestion chamber were found in the waste that left the digester. Predicted functions associated with degradation of xenobiotic compounds were significantly different between the sampling locations. The microbial community found in an anaerobic digestion reactor loaded with pig manure contained microorganisms with biochemical capacities related to the 4 phases of methane production. © 2017 S. Karger AG, Basel.

Understanding microbial ecology can help improve biogas production in AD.

PubMed

Ferguson, Robert M W; Coulon, Frédéric; Villa, Raffaella

2018-06-16

454-Pyrosequencing and lipid fingerprinting were used to link anaerobic digestion (AD) process parameters (pH, alkalinity, volatile fatty acids (VFAs), biogas production and methane content) with the reactor microbial community structure and composition. AD microbial communities underwent stress conditions after changes in organic loading rate and digestion substrates. 454-Pyrosequencing analysis showed that, irrespectively of the substrate digested, methane content and pH were always significantly, and positively, correlated with community evenness. In AD, microbial communities with more even distributions of diversity are able to use parallel metabolic pathways and have greater functional stability; hence, they are capable of adapting and responding to disturbances. In all reactors, a decrease in methane content to <30% was always correlated with a 50% increase of Firmicutes sequences (particularly in operational taxonomic units (OTUs) related to Ruminococcaceae and Veillonellaceae). Whereas digesters producing higher methane content (above 60%), contained a high number of sequences related to Synergistetes and unidentified bacterial OTUs. Finally, lipid fingerprinting demonstrated that, under stress, the decrease in archaeal biomass was higher than the bacterial one, and that archaeal Phospholipid etherlipids (PLEL) levels were correlated to reactor performances. These results demonstrate that, across a number of parameters such as lipids, alpha and beta diversity, and OTUs, knowledge of the microbial community structure can be used to predict, monitor, or optimise AD performance. Copyright © 2018 Elsevier B.V. All rights reserved.

Conceptual design of ACB-CP for ITER cryogenic system

NASA Astrophysics Data System (ADS)

Jiang, Yongcheng; Xiong, Lianyou; Peng, Nan; Tang, Jiancheng; Liu, Liqiang; Zhang, Liang

2012-06-01

ACB-CP (Auxiliary Cold Box for Cryopumps) is used to supply the cryopumps system with necessary cryogen in ITER (International Thermonuclear Experimental Reactor) cryogenic distribution system. The conceptual design of ACB-CP contains thermo-hydraulic analysis, 3D structure design and strength checking. Through the thermohydraulic analysis, the main specifications of process valves, pressure safety valves, pipes, heat exchangers can be decided. During the 3D structure design process, vacuum requirement, adiabatic requirement, assembly constraints and maintenance requirement have been considered to arrange the pipes, valves and other components. The strength checking has been performed to crosscheck if the 3D design meets the strength requirements for the ACB-CP.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1991-09-01

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 reactormore » 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.« less

Integrated head package for top mounted nuclear instrumentation

DOEpatents

Malandra, Louis J.; Hornak, Leonard P.; Meuschke, Robert E.

1993-01-01

A nuclear reactor such as a pressurized water reactor has an integrated head package providing structural support and increasing shielding leading toward the vessel head. A reactor vessel head engages the reactor vessel, and a control rod guide mechanism over the vessel head raises and lowers control rods in certain of the thimble tubes, traversing penetrations in the reactor vessel head, and being coupled to the control rods. An instrumentation tube structure includes instrumentation tubes with sensors movable into certain thimble tubes disposed in the fuel assemblies. Couplings for the sensors also traverse penetrations in the reactor vessel head. A shroud is attached over the reactor vessel head and encloses the control rod guide mechanism and at least a portion of the instrumentation tubes when retracted. The shroud forms a structural element of sufficient strength to support the vessel head, the control rod guide mechanism and the instrumentation tube structure, and includes radiation shielding material for limiting passage of radiation from retracted instrumentation tubes. The shroud is thicker at the bottom adjacent the vessel head, where the more irradiated lower ends of retracted sensors reside. The vessel head, shroud and contents thus can be removed from the reactor as a unit and rested safely and securely on a support.

Posttest Analyses of the Steel Containment Vessel Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Costello, J.F.; Hessheimer, M.F.; Ludwigsen, J.S.

A high pressure test of a scale model of a steel containment vessel (SCV) was conducted on December 11-12, 1996 at Sandia National Laboratories, Albuquerque, NM, USA. The test model is a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of an improved Mark II boiling water reactor (BWR) containment. This testis part of a program to investigate the response of representative models of nuclear containment structures to pressure loads beyond the design basis accident. The posttest analyses of this test focused on three areas where the pretest analysis effort did not adequately predict the model behavior duringmore » the test. These areas are the onset of global yielding, the strain concentrations around the equipment hatch and the strain concentrations that led to a small tear near a weld relief opening that was not modeled in the pretest analysis.« less

Laboratory Reactor for Processing Carbon-Containing Sludge

NASA Astrophysics Data System (ADS)

Korovin, I. O.; Medvedev, A. V.

2016-10-01

The paper describes a reactor for high-temperature pyrolysis of carbon-containing sludge with the possibility of further development of environmentally safe technology of hydrocarbon waste disposal to produce secondary products. A solution of the urgent problem has been found: prevention of environmental pollution resulting from oil pollution of soils using the pyrolysis process as a method of disposal of hydrocarbon waste to produce secondary products.

METHOD OF DEHYDRATING URANIUM TETRAFLUORIDE

DOEpatents

Davis, J.O.; Fogel, C.C.; Palmer, W.E.

1962-12-18

Drying and dehydration of aqueous-precipitated uranium tetrafluoride are described. The UF/sub 4/ which normally contains 3 to 4% water, is dispersed into the reaction zone of an operating reactor wherein uranium hexafluoride is being reduced to UF/sub 4/ with hydrogen. The water-containing UF/sub 4/ is dried and blended with the UF/sub 4/ produced in the reactor without interfering with the reduction reaction. (AEC)

Impact of non-ionic surfactant on the long-term development of lab-scale-activated sludge bacterial communities.

PubMed

Lozada, Mariana; Basile, Laura; Erijman, Leonardo

2007-01-01

The development of bacterial communities in replicate lab-scale-activated sludge reactors degrading a non-ionic surfactant was evaluated by statistical analysis of denaturing gradient gel electrophoresis (DGGE) fingerprints. Four sequential batch reactors were fed with synthetic sewage, two of which received, in addition, 0.01% of nonylphenol ethoxylates (NPE). The dynamic character of bacterial community structure was confirmed by the differences in species composition among replicate reactors. Measurement of similarities between reactors was obtained by pairwise similarity analysis using the Bray Curtis coefficient. The group of NPE-amended reactors exhibited the highest similarity values (Sjk=0.53+/-0.03), indicating that the bacterial community structure of NPE-amended reactors was better replicated than control reactors (Sjk=0.36+/-0.04). Replicate NPE-amended reactors taken at different times of operation clustered together, whereas analogous relations within the control reactor cluster were not observed. The DGGE pattern of isolates grown in conditioned media prepared with media taken at the end of the aeration cycle grouped separately from other conditioned and synthetic media regardless of the carbon source amendment, suggesting that NPE degradation residuals could have a role in the shaping of the community structure.

Reactor water cleanup system

DOEpatents

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

1994-12-20

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

Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

NASA Astrophysics Data System (ADS)

Seibert, Rachel L.; Terrani, Kurt A.; Velázquez, Daniel; Hunn, John D.; Baldwin, Charles A.; Montgomery, Fred C.; Terry, Jeff

2018-03-01

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Program and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form PdxSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. They may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.

Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

DOE PAGES

Seibert, Rachel L.; Terrani, Kurt A.; Velázquez, Daniel; ...

2018-03-01

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Programmore » and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form Pd xSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. In conclusion, they may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.« less

Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

DOE Office of Scientific and Technical Information (OSTI.GOV)

Seibert, Rachel L.; Terrani, Kurt A.; Velázquez, Daniel

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Programmore » and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form Pd xSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. In conclusion, they may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.« less

Anammox enrichment from reject water on blank biofilm carriers and carriers containing nitrifying biomass: operation of two moving bed biofilm reactors (MBBR).

PubMed

Zekker, Ivar; Rikmann, Ergo; Tenno, Toomas; Lemmiksoo, Vallo; Menert, Anne; Loorits, Liis; Vabamäe, Priit; Tomingas, Martin; Tenno, Taavo

2012-07-01

The anammox bacteria were enriched from reject water of anaerobic digestion of municipal wastewater sludge onto moving bed biofilm reactor (MBBR) system carriers-the ones initially containing no biomass (MBBR1) as well as the ones containing nitrifying biomass (MBBR2). Duration of start-up periods of the both reactors was similar (about 100 days), but stable total nitrogen (TN) removal efficiency occurred earlier in the system containing nitrifying biomass. Anammox TN removal efficiency of 70% was achieved by 180 days in both 20 l volume reactors at moderate temperature of 26.0°C. During the steady state phase of operation of MBBRs the average TN removal efficiencies and maximum TN removal rates in MBBR1 were 80% (1,000 g-N/m(3)/day, achieved by 308 days) and in MBBR2 85% (1,100 g-N/m(3)/day, achieved by 266 days). In both reactors mixed bacterial cultures were detected. Uncultured Planctomycetales bacterium clone P4, Candidatus Nitrospira defluvii and uncultured Nitrospira sp. clone 53 were identified by PCR-DGGE from the system initially containing blank biofilm carriers as well as from the nitrifying biofilm system; from the latter in addition to these also uncultured ammonium oxidizing bacterium clone W1 and Nitrospira sp. clone S1-62 were detected. FISH analysis revealed that anammox microorganisms were located in clusters in the biofilm. Using previously grown nitrifying biofilm matrix for anammox enrichment has some benefits over starting up the process from zero, such as less time for enrichment and protection against severe inhibitions in case of high substrate loading rates.

Nuclear reactor building

DOEpatents

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

1994-04-05

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

Nuclear reactor building

DOEpatents

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

1994-01-01

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

Method for continuously recovering metals using a dual zone chemical reactor

DOEpatents

Bronson, Mark C.

1995-01-01

A dual zone chemical reactor continuously processes metal-containing materials while regenerating and circulating a liquid carrier. The starting materials are fed into a first reaction zone of a vessel containing a molten salt carrier. The starting materials react to form a metal product and a by-product that dissolves in the molten salt that flows to a second reaction zone in the reaction vessel. The second reaction zone is partitioned from, but in fluid communication with, the first reaction zone. The liquid carrier continuously circulates along a pathway between the first reaction zone and the second reaction zone. A reactive gas is introduced into the second reaction zone to react with the reaction by-product to generate the molten salt. The metal product, the gaseous waste products, and the excess liquid carrier are removed without interrupting the operation of the reactor. The design of the dual zone reactor can be adapted to combine a plurality of liquid carrier regeneration zones in a multiple dual zone chemical reactor for production scale processing.

Fossil fuel furnace reactor

DOEpatents

Parkinson, William J.

1987-01-01

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

Changing concepts of geologic structure and the problem of siting nuclear reactors: Examples from Washington State

NASA Astrophysics Data System (ADS)

Tabor, R. W.

1986-09-01

The conflict between regulation and healthy evolution of geological science has contributed to the difficulties of siting nuclear reactors. On the Columbia Plateau in Washington, but for conservative design of the Hanford reactor facility, the recognition of the little-understood Olympic-Wallowa lineament as a major, possibly still active structural alinement might have jeopardized the acceptability of the site for nuclear reactors. On the Olympic Peninsula, evolving concepts of compressive structures and their possible recent activity and the current recognition of a subducting Juan de Fuca plate and its potential for generating great earthquakes—both concepts little-considered during initial site selection—may delay final acceptance of the Satsop site. Conflicts of this sort are inevitable but can be accommodated if they are anticipated in the reactor-licensing process. More important, society should be increasing its store of geologic knowledge now, during the current recess in nuclear reactor siting.

Solar-thermal fluid-wall reaction processing

DOEpatents

Weimer, Alan W.; Dahl, Jaimee K.; Lewandowski, Allan A.; Bingham, Carl; Buechler, Karen J.; Grothe, Willy

2006-04-25

The present invention provides a method for carrying out high temperature thermal dissociation reactions requiring rapid-heating and short residence times using solar energy. In particular, the present invention provides a method for carrying out high temperature thermal reactions such as dissociation of hydrocarbon containing gases and hydrogen sulfide to produce hydrogen and dry reforming of hydrocarbon containing gases with carbon dioxide. In the methods of the invention where hydrocarbon containing gases are dissociated, fine carbon black particles are also produced. The present invention also provides solar-thermal reactors and solar-thermal reactor systems.

Solar-Thermal Fluid-Wall Reaction Processing

DOEpatents

Weimer, A. W.; Dahl, J. K.; Lewandowski, A. A.; Bingham, C.; Raska Buechler, K. J.; Grothe, W.

2006-04-25

The present invention provides a method for carrying out high temperature thermal dissociation reactions requiring rapid-heating and short residence times using solar energy. In particular, the present invention provides a method for carrying out high temperature thermal reactions such as dissociation of hydrocarbon containing gases and hydrogen sulfide to produce hydrogen and dry reforming of hydrocarbon containing gases with carbon dioxide. In the methods of the invention where hydrocarbon containing gases are dissociated, fine carbon black particles are also produced. The present invention also provides solar-thermal reactors and solar-thermal reactor systems.

Skyshine radiation from a pressurized water reactor containment dome

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peng, W.H.

1986-06-01

The radiation dose rates resulting from airborne activities inside a postaccident pressurized water reactor containment are calculated by a discrete ordinates/Monte Carlo combined method. The calculated total dose rates and the skyshine component are presented as a function of distance from the containment at three different elevations for various gamma-ray source energies. The one-dimensional (ANISN code) is used to approximate the skyshine dose rates from the hemisphere dome, and the results are compared favorably to more rigorous results calculated by a three-dimensional Monte Carlo code.

Integrated production of fuel gas and oxygenated organic compounds from synthesis gas

DOEpatents

Moore, Robert B.; Hegarty, William P.; Studer, David W.; Tirados, Edward J.

1995-01-01

An oxygenated organic liquid product and a fuel gas are produced from a portion of synthesis gas comprising hydrogen, carbon monoxide, carbon dioxide, and sulfur-containing compounds in a integrated feed treatment and catalytic reaction system. To prevent catalyst poisoning, the sulfur-containing compounds in the reactor feed are absorbed in a liquid comprising the reactor product, and the resulting sulfur-containing liquid is regenerated by stripping with untreated synthesis gas from the reactor. Stripping offgas is combined with the remaining synthesis gas to provide a fuel gas product. A portion of the regenerated liquid is used as makeup to the absorber and the remainder is withdrawn as a liquid product. The method is particularly useful for integration with a combined cycle coal gasification system utilizing a gas turbine for electric power generation.

Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1993-09-15

This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

NUCLEAR REACTOR

DOEpatents

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

1961-06-20

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

METHOD AND APPARATUS FOR HANDLING RADIOACTIVE PRODUCTS

DOEpatents

Nicoll, D.

1959-02-24

A device is described for handling fuel elements being discharged from a nuclear reactor. The device is adapted to be disposed beneath a reactor within the storage canal for spent fuel elements. The device is comprised essentially of a cylinder pivotally mounted to a base for rotational motion between a vertical position. where the mouth of the cylinder is in the top portion of the container for receiving a fuel element discharged from a reactor into the cylinder, and a horizontal position where the mouth of the cylinder is remote from the top portion of the container and the fuel element is discharged from the cylinder into the storage canal. The device is operated by hydraulic pressure means and is provided with a means to prevent contaminated primary liquid coolant in the reactor system from entering the storage canal with the spent fuel element.

Process for photosynthetically splitting water

DOEpatents

Greenbaum, Elias

1984-01-01

The invention is an improved process for producing gaseous hydrogen and oxygen from water. The process is conducted in a photolytic reactor which contains a water-suspension of a photoactive material containing a hydrogen-liberating catalyst. The reactor also includes a volume for receiving gaseous hydrogen and oxygen evolved from the liquid phase. To avoid oxygen-inactivation of the catalyst, the reactor is evacuated continuously by an external pump which circulates the evolved gases through means for selectively recovering hydrogen therefrom. The pump also cools the reactor by evaporating water from the liquid phase. Preferably, product recovery is effected by selectively diffusing the hydrogen through a heated semipermeable membrane, while maintaining across the membrane a magnetic field gradient which biases the oxygen away from the heated membrane. This promotes separation, minimizes the back-reaction of hydrogen and oxygen, and protects the membrane.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-07-01

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

MELCOR computer code manuals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

New PANDA Tests to Investigate Effects of Light Gases on Passive Safety Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paladino, D.; Auban, O.; Candreia, P.

The large- scale thermal-hydraulic PANDA facility (located at PSI in Switzerland), has been used over the last few years for investigating different passive decay- heat removal systems and containment phenomena for the next generation of light water reactors (Simplified Boiling Water Reactor: SBWR; European Simplified Boiling Water Reactor: ESBWR; Siedewasserreaktor: SWR-1000). Currently, as part of the European Commission 5. EURATOM Framework Programme project 'Testing and Enhanced Modelling of Passive Evolutionary Systems Technology for Containment Cooling' (TEMPEST), a new series of tests is being planned in the PANDA facility to experimentally investigate the distribution of non-condensable gases inside the containment andmore » their effect on the performance of the 'Passive Containment Cooling System' (PCCS). Hydrogen release caused by the metal-water reaction in the case of a postulated severe accident will be simulated in PANDA by injecting helium into the reactor pressure vessel. In order to provide suitable data for Computational Fluid Dynamic (CFD) code assessment and improvement, the instrumentation in PANDA has been upgraded for the new tests. In the present paper, a detailed discussion is given of the new PANDA tests to be performed to investigate the effects of light gas on passive safety systems. The tests are scheduled for the first half of the year 2002. (authors)« less

Dual reactor for in situ/operando fluorescent mode XAS studies of sample containing low-concentration 3d or 5d metal elements

NASA Astrophysics Data System (ADS)

Nguyen, Luan; Tang, Yu; Li, Yuting; Zhang, Xiaoyan; Wang, Ding; Tao, Franklin Feng

2018-05-01

Transition metal elements are the most important elements of heterogeneous catalysts used for chemical and energy transformations. Many of these catalysts are active at a temperature higher than 400 °C. For a catalyst containing a 3d or 5d metal element with a low concentration, typically their released fluorescence upon the K-edge or L-edge adsorption of X-rays is collected for the analysis of chemical and coordination environments of these elements. However, it is challenging to perform in situ/operando X-ray absorption spectroscopy (XAS) studies of elements of low-energy absorption edges at a low concentration in a catalyst during catalysis at a temperature higher than about 450 °C. Here a unique reaction system consisting two reactors, called a dual reactor system, was designed for performing in situ or operando XAS studies of these elements of low-energy absorption edges in a catalyst at a low concentration during catalysis at a temperature higher than 450 °C in a fluorescent mode. This dual-reactor system contains a quartz reactor for preforming high-temperature catalysis up to 950 °C and a Kapton reactor remaining at a temperature up to 450 °C for collecting data in the same gas of catalysis. With this dual reactor, chemical and coordination environments of low-concentration metal elements with low-energy absorption edges such as the K-edge of 3d metals including Ti, V, Cr, Mn, Fe, Co, Ni, and Cu and L edge of 5d metals including W, Re, Os, Ir, Pt, and Au can be examined through first performing catalysis at a temperature higher than 450 °C in the quartz reactor and then immediately flipping the catalyst in the same gas flow to the Kapton reactor remained up to 450 °C to collect data. The capability of this dual reactor was demonstrated by tracking the Mn K-edge of the MnOx/Na2WO4 catalyst during activation in the temperature range of 300-900 °C and catalysis at 850 °C.

A comparative study of thermophilic and mesophilic anaerobic co-digestion of food waste and wheat straw: Process stability and microbial community structure shifts.

PubMed

Shi, Xuchuan; Guo, Xianglin; Zuo, Jiane; Wang, Yajiao; Zhang, Mengyu

2018-05-01

Renewable energy recovery from organic solid waste via anaerobic digestion is a promising way to provide sustainable energy supply and eliminate environmental pollution. However, poor efficiency and operational problems hinder its wide application of anaerobic digestion. The effects of two key parameters, i.e. temperature and substrate characteristics on process stability and microbial community structure were studied using two lab-scale anaerobic reactors under thermophilic and mesophilic conditions. Both the reactors were fed with food waste (FW) and wheat straw (WS). The organic loading rates (OLRs) were maintained at a constant level of 3 kg VS/(m 3 ·d). Five different FW:WS substrate ratios were utilized in different operational phases. The synergetic effects of co-digestion improved the stability and performance of the reactors. When FW was mono-digested, both reactors were unstable. The mesophilic reactor eventually failed due to volatile fatty acid accumulation. The thermophilic reactor had better performance compared to mesophilic one. The biogas production rate of the thermophilic reactor was 4.9-14.8% higher than that of mesophilic reactor throughout the experiment. The shifts in microbial community structures throughout the experiment in both thermophilic and mesophilic reactors were investigated. With increasing FW proportions, bacteria belonging to the phylum Thermotogae became predominant in the thermophilic reactor, while the phylum Bacteroidetes was predominant in the mesophilic reactor. The genus Methanosarcina was the predominant methanogen in the thermophilic reactor, while the genus Methanothrix remained predominant in the mesophilic reactor. The methanogenesis pathway shifted from acetoclastic to hydrogenotrophic when the mesophilic reactor experienced perturbations. Moreover, the population of lignocellulose-degrading microorganisms in the thermophilic reactor was higher than those in mesophilic reactor, which explained the better performance of the thermophilic reactor. Copyright © 2018. Published by Elsevier Ltd.

A coupled nuclear reactor thermal energy storage system for enhanced load following operation

NASA Astrophysics Data System (ADS)

Alameri, Saeed A.

Nuclear power plants usually provide base-load electric power and operate most economically at a constant power level. In an energy grid with a high fraction of renewable energy sources, future nuclear reactors may be subject to significantly variable power demands. These variable power demands can negatively impact the effective capacity factor of the reactor and result in severe economic penalties. Coupling the reactor to a large Thermal Energy Storage (TES) block will allow the reactor to better respond to variable power demands. In the system described in this thesis, a Prismatic-core Advanced High Temperature Reactor (PAHTR) operates at constant power with heat provided to a TES block that supplies power as needed to a secondary energy conversion system. The PAHTR is designed to have a power rating of 300 MW th, with 19.75 wt% enriched Tri-Structural-Isotropic UO 2 fuel and a five year operating cycle. The passive molten salt TES system will operate in the latent heat region with an energy storage capacity of 150 MWd. Multiple smaller TES blocks are used instead of one large block to enhance the efficiency and maintenance complexity of the system. A transient model of the coupled reactor/TES system is developed to study the behavior of the system in response to varying load demands. The model uses six-delayed group point kinetics and decay heat models coupled to thermal-hydraulic and heat transfer models of the reactor and TES system. Based on the transient results, the preferred TES design consists of 1000 blocks, each containing 11000 LiCl phase change material tubes. A safety assessment of major reactor events demonstrates the inherent safety of the coupled system. The loss of forced circulation study determined the minimum required air convection heat removal rate from the reactor core and the lowest possible reduced primary flow rate that can maintain the reactor in a safe condition. The loss of ultimate heat sink study demonstrated the ability of the TES to absorb the decay heat of the reactor fuel while cooling the PAHTR after an emergency shutdown. The simulated reactivity insertion accident assessment determined the maximum allowable reactivity insertion to the PAHTR as a function of shutdown response times.

FUEL ELEMENTS FOR NEUTRONIC REACTORS

DOEpatents

Foote, F.G.; Jette, E.R.

1963-05-01

A fuel element for a nuclear reactor is described that consists of a jacket containing a unitary core of fissionable material and a filling of a metal of the group consisting of sodium and sodium-potassium alloys. (AEC)

NEUTRONIC REACTOR

DOEpatents

Fermi, E.

1960-04-01

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

Electrowinning apparatus and process

DOEpatents

Buschmann, Wayne E [Boulder, CO

2012-06-19

Apparatus and processes are disclosed for electrowinning metal from a fluid stream. A representative apparatus comprises at least one spouted bed reactor wherein each said reactor includes an anolyte chamber comprising an anode and configured for containing an anolyte, a catholyte chamber comprising a current collector and configured for containing a particulate cathode bed and a flowing stream of an electrically conductive metal-containing fluid, and a membrane separating said anolyte chamber and said catholyte chamber, an inlet for an electrically conductive metal-containing fluid stream; and a particle bed churning device configured for spouting particle bed particles in the catholyte chamber independently of the flow of said metal-containing fluid stream. In operation, reduced heavy metals or their oxides are recovered from the cathode particles.

Calculation of the temperature in the container unit with a modified design for the production of {sup 99}Mo at the VVR-Ts research reactor facility (IVV.10M)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kazantsev, A. A., E-mail: kazantsevanatoly@gmail.com; Sergeev, V. V.; Kochnov, O. Yu.

The temperature regime is calculated for two different designs of containers with uranium-bearing material for the upgraded VVR-Ts research reactor facility (IVV.10M). The containers are to be used in the production of {sup 99}Mo. It is demonstrated that the modification of the container design leads to a considerable temperature reduction and an increase in the near-wall boiling margin and allows one to raise the amount of material loaded into the container. The calculations were conducted using the international thermohydraulic contour code TRAC intended to analyze the technical safety of water-cooled nuclear power units.

A modeling approach to describe ZVI-based anaerobic system.

PubMed

Xiao, Xiao; Sheng, Guo-Ping; Mu, Yang; Yu, Han-Qing

2013-10-15

Zero-valent iron (ZVI) is increasingly being added into anaerobic reactors to enhance the biological conversion of various less biodegradable pollutants (LBPs). Our study aimed to establish a new structure model based on the Anaerobic Digestion Model No. 1 (ADM1) to simulate such a ZVI-based anaerobic reactor. Three new processes, i.e., electron release from ZVI corrosion, H2 formation from ZVI corrosion, and transformation of LBPs, were integrated into ADM1. The established model was calibrated and tested using the experimental data from one published study, and validated using the data from another work. A good relationship between the predicted and measured results indicates that the proposed model was appropriate to describe the performance of the ZVI-based anaerobic system. Our model could provide more precise strategies for the design, development, and application of anaerobic systems especially for treating various LBPs-containing wastewaters. Copyright © 2013 Elsevier Ltd. All rights reserved.

Fatigue behavior of type 316 stainless steel following neutron irradiation inducing helium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grossbeck, M.L.; Liu, K.C.

1980-01-01

Since a tokamak fusion reactor operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially the first wall and blanket. Type 316 stainless steel in the 20% cold-worked condition has been irradiated in the HFIR in order to introduce helium as well as displacement damage. A miniature hourglass specimen was developed for the reactor irradiations and subsequent fully reversed low cycle fatigue testing. For material irradiated and tested at 430/sup 0/C in vacuum to a damage level of 7 to 15 dpa and containing 200 to 1000 appm He, a reduction in life by amore » factor of 3 to 10 was observed. An attempt was made to predict irradiated fatigue life by fitting data from irradiated material to a power law equation similar to the universal slopes equation and using ductility ratios from tensile tests to modify the equation for irradiated material.« less

Systems and methods for harvesting and storing materials produced in a nuclear reactor

DOEpatents

Heinold, Mark R.; Dayal, Yogeshwar; Brittingham, Martin W.

2016-04-05

Systems produce desired isotopes through irradiation in nuclear reactor instrumentation tubes and deposit the same in a robust facility for immediate shipping, handling, and/or consumption. Irradiation targets are inserted and removed through inaccessible areas without plant shutdown and placed in the harvesting facility, such as a plurality of sealable and shipping-safe casks and/or canisters. Systems may connect various structures in a sealed manner to avoid release of dangerous or unwanted matter throughout the nuclear plant, and/or systems may also automatically decontaminate materials to be released. Useable casks or canisters can include plural barriers for containment that are temporarily and selectively removable with specially-configured paths inserted therein. Penetrations in the facilities may limit waste or pneumatic gas escape and allow the same to be removed from the systems without over-pressurization or leakage. Methods include processing irradiation targets through such systems and securely delivering them in such harvesting facilities.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steiner, J.L.; Lime, J.F.; Elson, J.S.

One dimensional TRAC transient calculations of the process inherent ultimate safety (PIUS) advanced reactor design were performed for a pump-trip SCRAM. The TRAC calculations showed that the reactor power response and shutdown were in qualitative agreement with the one-dimensional analyses presented in the PIUS Preliminary Safety Information Document (PSID) submitted by Asea Brown Boveri (ABB) to the US Nuclear Regulatory Commission for preapplication safety review. The PSID analyses were performed with the ABB-developed RIGEL code. The TRAC-calculated phenomena and trends were also similar to those calculated with another one-dimensional PIUS model, the Brookhaven National Laboratory developed PIPA code. A TRACmore » pump-trip SCRAM transient has also been calculated with a TRAC model containing a multi-dimensional representation of the PIUS intemal flow structures and core region. The results obtained using the TRAC fully one-dimensional PIUS model are compared to the RIGEL, PIPA, and TRAC multi-dimensional results.« less

Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

NASA Astrophysics Data System (ADS)

Andrianova, E. A.; Tsibul'skiy, V. F.

2017-12-01

At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and 238U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with 238U alone should be reprocessed in the usual way.

Nuclear characteristics of a fissioning uranium plasma test reactor with light-water cooling

NASA Technical Reports Server (NTRS)

Whitmarsh, C. L., Jr.

1973-01-01

An analytical study was performed to determine a design configuration for a cavity test reactor. Test section criteria were that an average flux of 10 to the 15th power neutrons/sq cm/sec (E less than or equal to 0.12 eV) be supplied to a 61-cm-diameter spherical cavity at 200-atm pressure. Design objectives were to minimize required driver power, to use existing fuel-element technology, and to obtain fuel-element life of 10 to 100 full-power hours. Parameter calculations were made on moderator region size and material, driver fuel arrangement, control system, and structure in order to determine a feasible configuration. Although not optimized, a configuration was selected which would meet design criteria. The driver fuel region was a cylindrical annular region, one element thick, of 33 MTR-type H2O-cooled elements (Al-U fuel plate configuration), each 101 cm long. The region between the spherical test cavity and the cylindrical driver fuel region was Be (10 vol. % H2O coolant) with a midplane dimension of 8 cm. Exterior to the driver fuel, the 25-cm-thick cylindrical and axial reflectors were also Be with 10 vol. % H2O coolant. The entire reactor was contained in a 10-cm-thick steel pressure vessel, and the 200-atm cavity pressure was equalized throughout the driver reactor. Fuel-element life was 50 hr at the required driver power of 200 MW. Reactor control would be achieved with rotating poison drums located in the cylindrical reflector region. A control range of about 18 percent delta k/k was required for reactor operation.

NEUTRONIC REACTOR

DOEpatents

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

1958-09-01

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

Significance of breeding in fast nuclear reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Raza, S.M.; Abidi, S.B.M.

1983-12-01

Only breeder reactors--nuclear power plants that produce more fuel than they consume--are capable in principle of extracting the maximum amount of fission energy contained in uranium ore, thus offering a practical long-term solution to uranium supply problems. Uranium would then constitute a virtually inexhaustible fuel reserve for the world's future energy needs. The ultimate argument for breeding is to conserve the energy resources available to mankind. A long-term role for nuclear power with fast reactors is proven to be economically viable, environmentally acceptable and capable of wide scale exploitation in many countries. In this paper, various suggestions pertaining to themore » fuel fabrication route, fuel cycle economics, studies of the physics of fast nuclear reactors and of engineering design simplifications are presented. Fast reactors contain no moderator and inherently require enriched fuel. In general, the main aim is to suggest an improvement in the understanding of the safety and control characteristics of fast breeder power reactors. Development work is also being devoted to new carbide and nitride fuels, which are likely to exhibit breeding characteristics superior to those of the oxides of plutonium and uranium.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Merzari, E.; Yuan, Haomin; Kraus, A.

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable componentsmore » in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the design stage. Over the past five years, the Reactor Product Line has developed the integrated multi-physics code suite SHARP. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside the full reactor core or portions of it with a user-specified fidelity. In particular SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. These tools form a strong basis on which to build a flow-induced vibration modeling capability. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo aimed at simulating available FIV experiments in helical steam generators in the turbulent buffeting regime. In this regime one-way coupling is judged sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators at the low flows expected in integral PWRs. The legacy data is obtained from two datasets developed at Argonne and B&W.« less

Aseismic safety analysis of a prestressed concrete containment vessel for CPR1000 nuclear power plant

NASA Astrophysics Data System (ADS)

Yi, Ping; Wang, Qingkang; Kong, Xianjing

2017-01-01

The containment vessel of a nuclear power plant is the last barrier to prevent nuclear reactor radiation. Aseismic safety analysis is the key to appropriate containment vessel design. A prestressed concrete containment vessel (PCCV) model with a semi-infinite elastic foundation and practical arrangement of tendons has been established to analyze the aseismic ability of the CPR1000 PCCV structure under seismic loads and internal pressure. A method to model the prestressing tendon and its interaction with concrete was proposed and the axial force of the prestressing tendons showed that the simulation was reasonable and accurate. The numerical results show that for the concrete structure, the location of the cylinder wall bottom around the equipment hatch and near the ring beam are critical locations with large principal stress. The concrete cracks occurred at the bottom of the PCCV cylinder wall under the peak earthquake motion of 0.50 g, however the PCCV was still basically in an elastic state. Furthermore, the concrete cracks occurred around the equipment hatch under the design internal pressure of 0.4MPa, but the steel liner was still in the elastic stage and its leak-proof function soundness was verified. The results provide the basis for analysis and design of containment vessels.

Liquid phase methanol reactor staging process for the production of methanol

DOEpatents

Bonnell, Leo W.; Perka, Alan T.; Roberts, George W.

1988-01-01

The present invention is a process for the production of methanol from a syngas feed containing carbon monoxide, carbon dioxide and hydrogen. Basically, the process is the combination of two liquid phase methanol reactors into a staging process, such that each reactor is operated to favor a particular reaction mechanism. In the first reactor, the operation is controlled to favor the hydrogenation of carbon monoxide, and in the second reactor, the operation is controlled so as to favor the hydrogenation of carbon dioxide. This staging process results in substantial increases in methanol yield.

Thermos reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Labrousse, M.; Lerouge, B.; Dupuy, G.

1978-04-01

THERMOS is a water reactor designed to provide hot water up to 120/sup 0/C for district heating or for desalination applications. It is a 100-MW reactor based on proven technology: oxide fuel plate elements, integrated primary circuit, and reactor vessel located in the bottom of a pool. As in swimming pool reactors, the pool is used for biological shielding, emergency core cooling, and fission product filtering (in case of an accident). Before economics, safety is the main characteristic of the concept: no fuel failure admitted, core under water in any accidental configuration, inspection of every ''nuclear'' component, and double-wall containment.

Microfluidic electrochemical reactors

DOEpatents

Nuzzo, Ralph G [Champaign, IL; Mitrovski, Svetlana M [Urbana, IL

2011-03-22

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

Process of forming catalytic surfaces for wet oxidation reactions

NASA Technical Reports Server (NTRS)

Jagow, R. B. (Inventor)

1977-01-01

A wet oxidation process was developed for oxidizing waste materials, comprising dissolved ruthenium salt in a reactant feed stream containing the waste materials. The feed stream is introduced into a reactor, and the reactor contents are then raised to an elevated temperature to effect deposition of a catalytic surface of ruthenium black on the interior walls of the reactor. The feed stream is then maintained in the reactor for a period of time sufficient to effect at least partial oxidation of the waste materials.

DENSITY CONTROL IN A REACTOR

DOEpatents

Marshall, J. Jr.

1961-10-24

A reactor is described in which natural-uranium bodies are located in parallel channels which extend through the graphite mass in a regular lattice. The graphite mass has additional channels that are out of the lattice and contain no uranium. These additional channels decrease in number per unit volume of graphite from the center of the reactor to the exterior and have the effect of reducing the density of the graphite more at the center than at the exterior, thereby spreading neutron activity throughout the reactor. (AEC)

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.

1983-06-01

During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

Fuel assembly for the production of tritium in light water reactors

DOEpatents

Cawley, W.E.; Trapp, T.J.

1983-06-10

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

Fuel assembly for the production of tritium in light water reactors

DOEpatents

Cawley, William E.; Trapp, Turner J.

1985-01-01

A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

Finite Element Stress Analysis of Spent Nuclear Fuel Disposal Canister in a Deep Geological Repository

NASA Astrophysics Data System (ADS)

Kwon, Young Joo; Choi, Jong Won

This paper presents the finite element stress analysis of a spent nuclear fuel disposal canister to provide basic information for dimensioning the canister and configuration of canister components and consequently to suggest the structural analysis methodology for the disposal canister in a deep geological repository which is nowadays very important in the environmental waste treatment technology. Because of big differences in the pressurized water reactor (PWR) and the Canadian deuterium and uranium reactor (CANDU) fuel properties, two types of canisters are conceived. For manufacturing, operational reasons and standardization, however, both canisters have the same outer diameter and length. The construction type of canisters introduced here is a solid structure with a cast insert and a corrosion resistant overpack. The structural stress analysis is carried out using a finite element analysis code, NISA, and focused on the structural strength of the canister against the expected external pressures due to the swelling of the bentonite buffer and the hydrostatic head. The canister must withstand these large pressure loads. Consequently, canisters presented here contain 4 PWR fuel assemblies and 33×9 CANDU fuel bundles. The outside diameter of the canister for both fuels is 122cm and the cast insert diameter is 112cm. The total length of the canister is 483cm with the lid/bottom and the outer shell of 5cm.

Post impact behavior of mobile reactor core containment systems

NASA Technical Reports Server (NTRS)

Puthoff, R. L.; Parker, W. G.; Vanbibber, L. E.

1972-01-01

The reactor core containment vessel temperatures after impact, and the design variables that affect the post impact survival of the system are analyzed. The heat transfer analysis includes conduction, radiation, and convection in addition to the core material heats of fusion and vaporization under partially burial conditions. Also, included is the fact that fission products vaporize and transport radially outward and condense outward and condense on cooler surfaces, resulting in a moving heat source. A computer program entitled Executive Subroutines for Afterheat Temperature Analysis (ESATA) was written to consider this complex heat transfer analysis. Seven cases were calculated of a reactor power system capable of delivering up to 300 MW of thermal power to a nuclear airplane.

Spherical torus fusion reactor

DOEpatents

Peng, Yueng-Kay M.

1989-04-04

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

Spherical torus fusion reactor

DOEpatents

Peng, Yueng-Kay M.

1989-01-01

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

Fission product transport analysis in a loss of decay heat removal accident at Browns Ferry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wichner, R.P.; Weber, C.F.; Hodge, S.A.

1984-01-01

This paper summarizes an analysis of the movement of noble gases, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal (DHR) capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris ontomore » the drywell floor.« less

Internal combustion engine having a reactor for afterburning of unburned exhaust gas constituents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maurhoff, G.; Steinwart, J.

1974-08-07

An internal combustion engine is described which has an engine housing and a reactor for afterburning of unburned constituents in the exhaust gas. The reactor has a shell with a periphery and contains a heat-insulated, reactor chamber which is freely movable beyond the point of connection to the shell. The reactor has an inlet nozzle extending freely through the shell and connected to an outlet passage of the engine and has an outlet for escape of the exhaust gases from the reactor chamber. The inlet nozzle protrudes freely into the outlet passage, and the shell has a portion around themore » inlet nozzle in contact with the engine housing.« less

Mechanical design of a light water breeder reactor

DOEpatents

Fauth, Jr., William L.; Jones, Daniel S.; Kolsun, George J.; Erbes, John G.; Brennan, John J.; Weissburg, James A.; Sharbaugh, John E.

1976-01-01

In a light water reactor system using the thorium-232 -- uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements.

SODIUM DEUTERIUM REACTOR

DOEpatents

Oppenheimer, E.D.; Weisberg, R.A.

1963-02-26

This patent relates to a barrier system for a sodium heavy water reactor capable of insuring absolute separation of the metal and water. Relatively cold D/sub 2/O moderator and reflector is contained in a calandria into which is immersed the fuel containing tubes. The fuel elements are cooled by the sodium which flows within the tubes and surrounds the fuel elements. The fuel containing tubes are surrounded by concentric barrier tubes forming annular spaces through which pass inert gases at substantially atmospheric pressure. Header rooms above and below the calandria are provided for supplying and withdrawing the sodium and inert gases in the calandria region. (AEC)

Single Crystal Diffuse Neutron Scattering

DOE PAGES

Welberry, Richard; Whitfield, Ross

2018-01-11

Diffuse neutron scattering has become a valuable tool for investigating local structure in materials ranging from organic molecular crystals containing only light atoms to piezo-ceramics that frequently contain heavy elements. Although neutron sources will never be able to compete with X-rays in terms of the available flux the special properties of neutrons, viz. the ability to explore inelastic scattering events, the fact that scattering lengths do not vary systematically with atomic number and their ability to scatter from magnetic moments, provides strong motivation for developing neutron diffuse scattering methods. Here, we compare three different instruments that have been used bymore » us to collect neutron diffuse scattering data. Two of these are on a spallation source and one on a reactor source.« less

Single Crystal Diffuse Neutron Scattering

DOE Office of Scientific and Technical Information (OSTI.GOV)

Welberry, Richard; Whitfield, Ross

Diffuse neutron scattering has become a valuable tool for investigating local structure in materials ranging from organic molecular crystals containing only light atoms to piezo-ceramics that frequently contain heavy elements. Although neutron sources will never be able to compete with X-rays in terms of the available flux the special properties of neutrons, viz. the ability to explore inelastic scattering events, the fact that scattering lengths do not vary systematically with atomic number and their ability to scatter from magnetic moments, provides strong motivation for developing neutron diffuse scattering methods. Here, we compare three different instruments that have been used bymore » us to collect neutron diffuse scattering data. Two of these are on a spallation source and one on a reactor source.« less

Qualification of data obtained during a severe accident. Illustrative examples from TMI-2 evaluations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rempe, Joy L.; Knudson, Darrell L.

2015-02-01

The accidents at the Three Mile Island Unit 2 (TMI-2) Pressurized Water Reactor (PWR) and the Daiichi Units 1, 2, and 3 Boiling Water Reactors (BWRs) provide unique opportunities to evaluate instrumentation exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during the TMI-2 accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. Post-TMI-2 instrumentation evaluation programs focused on data required by TMI-2 operators to assess the condition of the reactor and containment and the effect of mitigating actions taken bymore » these operators. Prior efforts also focused on sensors providing data required for subsequent forensic evaluations and accident simulations. This paper provides additional details related to the formal process used to develop a qualified TMI-2 data base and presents data qualification details for three parameters: reactor coolant system (RCS) pressure; containment building temperature; and containment pressure. These selected examples illustrate the types of activities completed in the TMI-2 data qualification process and the importance of such a qualification effort. These details are described to facilitate implementation of a similar process using data and examinations at the Daiichi Units 1, 2, and 3 reactors so that BWR-specific benefits can be obtained.« less

Characterizing the biotransformation of sulfur-containing wastes in simulated landfill reactors.

PubMed

Sun, Wenjie; Sun, Mei; Barlaz, Morton A

2016-07-01

Landfills that accept municipal solid waste (MSW) in the U.S. may also accept a number of sulfur-containing wastes including residues from coal or MSW combustion, and construction and demolition (C&D) waste. Under anaerobic conditions that dominate landfills, microbially mediated processes can convert sulfate to hydrogen sulfide (H2S). The presence of H2S in landfill gas is problematic for several reasons including its low odor threshold, human toxicity, and corrosive nature. The objective of this study was to develop and demonstrate a laboratory-scale reactor method to measure the H2S production potential of a range of sulfur-containing wastes. The H2S production potential was measured in 8-L reactors that were filled with a mixture of the target waste, newsprint as a source of organic carbon required for microbial sulfate reduction, and leachate from decomposed residential MSW as an inoculum. Reactors were operated with and without N2 sparging through the reactors, which was designed to reduce H2S accumulation and toxicity. Both H2S and CH4 yields were consistently higher in reactors that were sparged with N2 although the magnitude of the effect varied. The laboratory-measured first order decay rate constants for H2S and CH4 production were used to estimate constants that were applicable in landfills. The estimated constants ranged from 0.11yr(-1) for C&D fines to 0.38yr(-1) for a mixed fly ash and bottom ash from MSW combustion. Copyright © 2016 Elsevier Ltd. All rights reserved.

Treatment of petroleum refinery wastewater containing heavily polluting substances in an aerobic submerged fixed-bed reactor.

PubMed

Vendramel, S; Bassin, J P; Dezotti, M; Sant'Anna, G L

2015-01-01

Petroleum refineries produce large amount of wastewaters, which often contain a wide range of different compounds. Some of these constituents may be recalcitrant and therefore difficult to be treated biologically. This study evaluated the capability of an aerobic submerged fixed-bed reactor (ASFBR) containing a corrugated PVC support material for biofilm attachment to treat a complex and high-strength organic wastewater coming from a petroleum refinery. The reactor operation was divided into five experimental runs which lasted more than 250 days. During the reactor operation, the applied volumetric organic load was varied within the range of 0.5-2.4 kgCOD.m(-3).d(-1). Despite the inherent fluctuations on the characteristics of the complex wastewater and the slight decrease in the reactor performance when the influent organic load was increased, the ASFBR showed good stability and allowed to reach chemical oxygen demand, dissolved organic carbon and total suspended solids removals up to 91%, 90% and 92%, respectively. Appreciable ammonium removal was obtained (around 90%). Some challenging aspects of reactor operation such as biofilm quantification and important biofilm constituents (e.g. polysaccharides (PS) and proteins (PT)) were also addressed in this work. Average PS/volatile attached solids (VAS) and PT/VAS ratios were around 6% and 50%, respectively. The support material promoted biofilm attachment without appreciable loss of solids and allowed long-term operation without clogging. Microscopic observations of the microbial community revealed great diversity of higher organisms, such as protozoa and rotifers, suggesting that toxic compounds found in the wastewater were possibly removed in the biofilm.

High-rate treatment of molasses wastewater by combination of an acidification reactor and a USSB reactor.

PubMed

Onodera, Takashi; Sase, Shinya; Choeisai, Pairaya; Yoochatchaval, Wilasinee; Sumino, Haruhiko; Yamaguchi, Takashi; Ebie, Yoshitaka; Xu, Kaiqin; Tomioka, Noriko; Syutsubo, Kazuaki

2011-01-01

A combination of an acidification reactor and an up-flow staged sludge bed (USSB) reactor was applied for treatment of molasses wastewater containing a large amount of organic compounds and sulfate. The USSB reactor had three gas-solid separators (GSS) along the height of the reactor. The combined system was continuously operated at mesophilic temperature over 400 days. In the acidification reactor, acid formation and sulfate reduction were effectively carried out. The sugars contained in the influent wastewater were mostly acidified into acetate, propionate, and n-butyrate. In addition, 10-30% of influent sulfur was removed from the acidification reactor by means of sulfate reduction followed by stripping of hydrogen sulfide. The USSB achieved a high organic loading rate (OLR) of 30 kgCOD m(-3) day(-1) with 82% COD removal. Vigorous biogas production was observed at a rate of 15 Nm(3) biogas m(-3) reactor day(-1). The produced biogas, including hydrogen sulfide, was removed from the wastewater mostly via the GSS. The GSS provided a moderate superficial biogas flux and low sulfide concentration in the sludge bed, resulting in the prevention of sludge washout and sulfide inhibition of methanogens. By advantages of this feature, the USSB may have been responsible for achieving sufficient retention (approximately 60 gVSS L(-1)) of the granular sludge with high methanogenic activity (0.88 gCOD gVSS(-1) day(-1) for acetate and as high as 2.6 gCOD gVSS(-1) day(-1) for H(2)/CO(2)). Analysis of the microbial community revealed that sugar-degrading acid-forming bacteria proliferated in the sludge of the USSB as well as the acidification reactor at high OLR conditions.

PRELIMINARY HAZARDS SUMMARY REPORT FOR THE VALLECITOS SUPERHEAT REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murray, J.L.

1961-02-01

BS>The Vallecitos Superheat Reactor (VSR) is a light-watermoderated, thermal-spectrum reactor, cooled by a combination of moderator boiling and forced convection cooling with saturated steam. The reactor core consists of 32 fuel hurdles containing 5300 lb of UO/sub 2/ enriched in U/sub 235/ to 3.6%. The fuel elements are arranged in individual process tubes that direct the cooling steam flow and separate the steam from the water moderator. The reactor vessel is designed for 1250 psig and operates at 960 to 1000 psig. With the reactor operating at 12.5 Mw(t), the maximum fuel cladding temperature is 1250 deg F and themore » cooling steam is superheated to an average temperature of about 810 deg F at 905 psig. Nu clear operation of the reactor is controlled by 12 control rods, actuated by drives mounted on the bottom of the reactor vessel. The water moderator recirculates inside the reactor vessel and through the core region by natural convection. Inherent safety features of the reactor include the negative core reactivity effects upon heating the UO/sub 2/ fuel (Doppler effect), upon increasing the temperature or void content of the moderator in the operating condition, and upon unflooding the fuel process tubes in the hot condition. Snfety features designed into the reactor and plant systems include a system of sensors and devices to detect petentially unsafe operating conditions and to initiate automatically the appropriate countermeasures, a set of fast and reliable control rods for scramming the reactor if a potentially unsafe condition occurs, a manually-actuated liquid neutron poison system, and an emergency cooling system to provide continued steam flow through the reactor core in the event the reactor becomes isolated from either its normal source of steam supply or discharge. The release of radioactivity to unrestricted areas is maintained within permissible limits by monitoring the radioactivity of wastes and controlling their release. The reactor and many of its auxiliaries are housed within a high-integrity essentially leak-tight containment vessel. (auth)« less

Pressurized reactor system and a method of operating the same

DOEpatents

Isaksson, J.M.

1996-06-18

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.

Pressurized reactor system and a method of operating the same

DOEpatents

Isaksson, Juhani M.

1996-01-01

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.

Interior of the Plum Brook Reactor Facility

NASA Image and Video Library

1961-02-21

A view inside the 55-foot high containment vessel of the National Aeronautics and Space Administration (NASA) Plum Brook Reactor Facility in Sandusky, Ohio. The 60-megawatt test reactor went critical for the first time in 1961 and began its full-power research operations in 1963. From 1961 to 1973, this reactor performed some of the nation’s most advanced nuclear research. The reactor was designed to determine the behavior of metals and other materials after long durations of irradiation. The materials would be used to construct a nuclear-powered rocket. The reactor core, where the chain reaction occurred, sat at the bottom of the tubular pressure vessel, seen here at the center of the shielding pool. The core contained fuel rods with uranium isotopes. A cooling system was needed to reduce the heat levels during the reaction. A neutron-impervious reflector was also employed to send many of the neutrons back to the core. The Plum Brook Reactor Facility was constructed from high-density concrete and steel to prevent the excess neutrons from escaping the facility, but the water in the pool shielded most of the radiation. The water, found in three of the four quadrants served as a reflector, moderator, and coolant. In this photograph, the three 20-ton protective shrapnel shields and hatch have been removed from the top of the pressure tank revealing the reactor tank. An overhead crane could be manipulated to reach any section of this room. It was used to remove the shrapnel shields and transfer equipment.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Picklesimer, M.L.; Thurber, W.C.

1961-01-01

A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

Method for continuously recovering metals using a dual zone chemical reactor

DOEpatents

Bronson, M.C.

1995-02-14

A dual zone chemical reactor continuously processes metal-containing materials while regenerating and circulating a liquid carrier. The starting materials are fed into a first reaction zone of a vessel containing a molten salt carrier. The starting materials react to form a metal product and a by-product that dissolves in the molten salt that flows to a second reaction zone in the reaction vessel. The second reaction zone is partitioned from, but in fluid communication with, the first reaction zone. The liquid carrier continuously circulates along a pathway between the first reaction zone and the second reaction zone. A reactive gas is introduced into the second reaction zone to react with the reaction by-product to generate the molten salt. The metal product, the gaseous waste products, and the excess liquid carrier are removed without interrupting the operation of the reactor. The design of the dual zone reactor can be adapted to combine a plurality of liquid carrier regeneration zones in a multiple dual zone chemical reactor for production scale processing. 6 figs.

Performance evaluation and phylogenetic characterization of anaerobic fluidized bed reactors using ground tire and pet as support materials for biohydrogen production.

PubMed

Barros, Aruana Rocha; Adorno, Maria Angela Tallarico; Sakamoto, Isabel Kimiko; Maintinguer, Sandra Imaculada; Varesche, Maria Bernadete Amâncio; Silva, Edson Luiz

2011-02-01

This study evaluated two different support materials (ground tire and polyethylene terephthalate [PET]) for biohydrogen production in an anaerobic fluidized bed reactor (AFBR) treating synthetic wastewater containing glucose (4000 mg L(-1)). The AFBR, which contained either ground tire (R1) or PET (R2) as support materials, were inoculated with thermally pretreated anaerobic sludge and operated at a temperature of 30°C. The AFBR were operated with a range of hydraulic retention times (HRT) between 1 and 8h. The reactor R1 operating with a HRT of 2h showed better performance than reactor R2, reaching a maximum hydrogen yield of 2.25 mol H(2)mol(-1) glucose with 1.3mg of biomass (as the total volatile solids) attached to each gram of ground tire. Subsequent 16S rRNA gene sequencing and phylogenetic analysis of particle samples revealed that reactor R1 favored the presence of hydrogen-producing bacteria such as Clostridium, Bacillus, and Enterobacter. Copyright © 2010 Elsevier Ltd. All rights reserved.

Microbial Community Structure and Diversity in an Integrated System of Anaerobic-Aerobic Reactors and a Constructed Wetland for the Treatment of Tannery Wastewater in Modjo, Ethiopia

PubMed Central

Desta, Adey Feleke; Assefa, Fassil; Leta, Seyoum; Stomeo, Francesca; Wamalwa, Mark; Njahira, Moses; Appolinaire, Djikeng

2014-01-01

A culture-independent approach was used to elucidate the microbial diversity and structure in the anaerobic-aerobic reactors integrated with a constructed wetland for the treatment of tannery wastewater in Modjo town, Ethiopia. The system has been running with removal efficiencies ranging from 94%–96% for COD, 91%–100% for SO42- and S2-, 92%–94% for BOD, 56%–82% for total Nitrogen and 2%–90% for NH3-N. 16S rRNA gene clone libraries were constructed and microbial community assemblies were determined by analysis of a total of 801 unique clone sequences from all the sites. Operational Taxonomic Unit (OTU) - based analysis of the sequences revealed highly diverse communities in each of the reactors and the constructed wetland. A total of 32 phylotypes were identified with the dominant members affiliated to Clostridia (33%), Betaproteobacteria (10%), Bacteroidia (10%), Deltaproteobacteria (9%) and Gammaproteobacteria (6%). Sequences affiliated to the class Clostridia were the most abundant across all sites. The 801 sequences were assigned to 255 OTUs, of which 3 OTUs were shared among the clone libraries from all sites. The shared OTUs comprised 80 sequences belonging to Clostridiales Family XIII Incertae Sedis, Bacteroidetes and unclassified bacterial group. Significantly different communities were harbored by the anaerobic, aerobic and rhizosphere sites of the constructed wetland. Numerous representative genera of the dominant bacterial classes obtained from the different sample sites of the integrated system have been implicated in the removal of various carbon- containing pollutants of natural and synthetic origins. To our knowledge, this is the first report of microbial community structure in tannery wastewater treatment plant from Ethiopia. PMID:25541981

Containment Sodium Chemistry Models in MELCOR.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Louie, David; Humphries, Larry L.; Denman, Matthew R

To meet regulatory needs for sodium fast reactors’ future development, including licensing requirements, Sandia National Laboratories is modernizing MELCOR, a severe accident analysis computer code developed for the U.S. Nuclear Regulatory Commission (NRC). Specifically, Sandia is modernizing MELCOR to include the capability to model sodium reactors. However, Sandia’s modernization effort primarily focuses on the containment response aspects of the sodium reactor accidents. Sandia began modernizing MELCOR in 2013 to allow a sodium coolant, rather than water, for conventional light water reactors. In the past three years, Sandia has been implementing the sodium chemistry containment models in CONTAIN-LMR, a legacy NRCmore » code, into MELCOR. These chemistry models include spray fire, pool fire and atmosphere chemistry models. Only the first two chemistry models have been implemented though it is intended to implement all these models into MELCOR. A new package called “NAC” has been created to manage the sodium chemistry model more efficiently. In 2017 Sandia began validating the implemented models in MELCOR by simulating available experiments. The CONTAIN-LMR sodium models include sodium atmosphere chemistry and sodium-concrete interaction models. This paper presents sodium property models, the implemented models, implementation issues, and a path towards validation against existing experimental data.« less

Investigation of Containment Flooding Strategy for Mark-III Nuclear Power Plant with MAAP4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Su Weinian; Wang, S.-J.; Chiang, S.-C

2005-06-15

Containment flooding is an important strategy for severe accident management of a conventional boiling water reactor (BWR) system. The purpose of this work is to investigate the containment flooding strategy of the Mark-III system after a reactor pressure vessel (RPV) breach. The Kuosheng Power Plant is a typical BWR-6 nuclear power plant (NPP) with Mark-III containment. The Severe Accident Management Guideline (SAMG) of the Kuosheng NPP has been developed based on the BWR Owners Group (BWROG) Emergency Procedure and Severe Accident Guidelines, Rev. 2. Therefore, the Kuosheng NPP is selected as the plant for study, and the MAAP4 code ismore » chosen as the tool for analysis. A postulated specific station blackout sequence for the Kuosheng NPP is cited as a reference case for this analysis. Because of the design features of Mark-III containment, the debris in the reactor cavity may not be submerged after an RPV breach when one follows the containment flooding strategy as suggested in the BWROG generic guideline, and the containment integrity could be challenged eventually. A more specific containment flooding strategy with drywell venting after an RPV breach is investigated, and a more stable plant condition is achieved with this strategy. Accordingly, the containment flooding strategy after an RPV breach will be modified for the Kuosheng SAMG, and these results are applicable to typical Mark-III plants with drywell vent path.« less

Effect of thiosulfate on rapid start-up of sulfur-based reduction of high concentrated perchlorate: A study of kinetics, extracellular polymeric substances (EPS) and bacterial community structure.

PubMed

Guo, Jianbo; Zhang, Chao; Lian, Jing; Lu, Caicai; Chen, Zhi; Song, Yuanyuan; Guo, Yankai; Xing, Yajuan

2017-11-01

Perchlorate (ClO 4 - ) contamination is more and more concerned due to the hazards to humans. Based on the common primary bacterium (Helicobacteraceae) of both thiosulfate-acclimated sludge (T-Acc) and sulfur-acclimated sludge (S-Acc) for perchlorate reduction, the rapid start-up of sulfur-based perchlorate reduction reactor (SBPRR) was hypothesized by inoculating T-Acc. Furthermore, the performance of SBPRR, the SO 4 2- yield, kinetics of ClO 4 - reduction and the extracellular polymeric substances (EPS) of biofilm confirmed the hypothesis. The start-up time of R3 (reactor inoculating T-Acc) was 0.18 and 0.21 times that of R1 (control) and R2 (reactor with the influent containing thiosulfate), respectively. The SO 4 2- yield of R3 was lower than that of R2 and R1 with perchlorate removal rate 166.7mg/(Lh). The kinetic study and EPS demonstrated that inoculating T-Acc was beneficial for the development of biofilm. Consequently, the present study indicated that SBPRR can be rapidly and successfully started-up via inoculation of T-Acc. Copyright © 2017 Elsevier Ltd. All rights reserved.

Experimental study on beryllium-7 production via sequential reactions in lithium-containing compounds irradiated by 14 MeV neutrons

NASA Astrophysics Data System (ADS)

Maekawa, F.; Verzilov, Y. M.; Smith, D. L.; Ikeda, Y.

2000-12-01

Except for 3H and 14C, no radioactive nuclide is produced by neutron-induced reactions with lithium in lithium-containing materials such as Li 2O and Li 2CO 3. However, when the lithium-containing materials are irradiated by 14 MeV neutrons, radioactive 7Be is produced by sequential charged particle reactions (SCPR). In this study, we measured effective 7Be production cross-sections in several lithium-containing samples at 14 MeV: the cross-sections are in the order of μb. Estimation of the effective cross-sections is attempted, and the estimated values agreed well with the experimental data. It was shown that the 7Be activity in a unit volume of lithium-containing materials in D-T fusion reactors can exceed total activity of the same unit volume of the SiC structural material in a certain cooling time. Consequently, a careful consideration of the 7Be production by SCPR is required to assess radioactive inventories in lithium-containing D-T fusion blanket materials.

LOFT. Containment and service building (TAN650). Section through north/south axis. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

LOFT. Containment and service building (TAN-650). Section through north/south axis. Shows basement and four additional levels of pre-amp tower, shielded roadway, chambers below reactor floor, railroad door, sumps, shielding. Section C shows basement sumps and chambers below reactor floor. Kaiser engineers 6413-11-STEP/LOFT-650-A-5. Date: October 1964. INEEL index code no. 036-650-00-486-122217 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

MELCOR computer code manuals: Primer and user`s guides, Version 1.8.3 September 1994. Volume 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users` Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

The combined effects of carbon/nitrogen ratio, suspended biomass, hydraulic retention time and dissolved oxygen on nutrient removal in a laboratory-scale anaerobic-anoxic-oxic activated sludge biofilm reactor.

PubMed

Manu, D S; Kumar Thalla, Arun

2018-01-01

The current trend in sustainable development deals mainly with environmental management. There is a need for economically affordable, advanced treatment methods for the proper treatment and management of domestic wastewater containing excess nutrients (such as nitrogen and phosphorus) which can cause eutrophication. The reduction of the excess nutrient content of wastewater by appropriate technology is of much concern to the environmentalist. In the current study, a novel integrated anaerobic-anoxic-oxic activated sludge biofilm (A 2 O-AS-biofilm) reactor was designed and operated to improve the biological nutrient removal by varying reactor operating conditions such as carbon to nitrogen (C/N) ratio, suspended biomass, hydraulic retention time (HRT) and dissolved oxygen (DO). Based on various trials, it was seen that the A 2 O-AS-biofilm reactor achieved good removal efficiencies with regard to chemical oxygen demand (95.5%), total phosphorus (93.1%), ammonia nitrogen concentration (NH 4 + -N) (98%) and total nitrogen (80%) when the reactor was maintained at C/N ratio of 4, suspended biomass of 3 to 3.5 g/L, HRT of 10 h, and DO of 1.5 to 2.5 mg/L. Scanning electron microscopy (SEM) of suspended and attached biofilm showed a dense structure of coccus and bacillus bacteria with the diameter ranging from 0.3 to 1.2 μm. The Fourier transform infrared (FTIR) spectroscopy results indicated phosphorylated macromolecules and carbohydrates mix or bind with extracellular proteins in exopolysaccharides.

Pentachlorophenol (PCP) dechlorination in horizontal-flow anaerobic immobilized biomass (HAIB) reactors.

PubMed

Damianovic, M H R Z; Moraes, E M; Zaiat, M; Foresti, E

2009-10-01

This study verifies the potential applicability of horizontal-flow anaerobic immobilized biomass (HAIB) reactors to pentachlorophenol (PCP) dechlorination. Two bench-scale HAIB reactors (R1 and R2) were filled with cubic polyurethane foam matrices containing immobilized anaerobic sludge. The reactors were then continuously fed with synthetic wastewater consisting of PCP, glucose, acetic acid, and formic acid as co-substrates for PCP anaerobic degradation. Before being immobilized in polyurethane foam matrices, the biomass was exposed to wastewater containing PCP in reactors fed at a semi-continuous rate of 2.0 microg PCP g(-1) VS. The applied PCP loading rate was increased from 0.05 to 2.59 mg PCP l(-1)day(-1) for R1, and from 0.06 to 4.15 mg PCP l(-1)day(-1) for R2. The organic loading rates (OLR) were 1.1 and 1.7 kg COD m(-3)day(-1) at hydraulic retention times (HRT) of 24h for R1 and 18 h for R2. Under such conditions, chemical oxygen demand (COD) removal efficiencies of up to 98% were achieved in the HAIB reactors. Both reactors exhibited the ability to remove 97% of the loaded PCP. Dichlorophenol (DCP) was the primary chlorophenol detected in the effluent. The adsorption of PCP and metabolites formed during PCP degradation in the packed bed was negligible for PCP removal efficiency.

BOILING SLURRY REACTOR AND METHOD FO CONTROL

DOEpatents

Petrick, M.; Marchaterre, J.F.

1963-05-01

The control of a boiling slurry nuclear reactor is described. The reactor consists of a vertical tube having an enlarged portion, a steam drum at the top of the vertical tube, and at least one downcomer connecting the steam drum and the bottom of the vertical tube, the reactor being filled with a slurry of fissionabie material in water of such concentration that the enlarged portion of the vertical tube contains a critical mass. The slurry boils in the vertical tube and circulates upwardly therein and downwardly in the downcomer. To control the reactor by controlling the circulation of the slurry, a gas is introduced into the downcomer. (AEC)

Exhaust system with emissions storage device and plasma reactor

DOEpatents

Hoard, John W.

1998-01-01

An exhaust system for a combustion system, comprising a storage device for collecting NO.sub.x, hydrocarbon, or particulate emissions, or mixture of these emissions, and a plasma reactor for destroying the collected emissions is described. After the emission is collected in by the storage device for a period of time, the emission is then destroyed in a non-thermal plasma generated by the plasma reactor. With respect to the direction of flow of the exhaust stream, the storage device must be located before the terminus of the plasma reactor, and it may be located wholly before, overlap with, or be contained within the plasma reactor.

Cavity temperature and flow characteristics in a gas-core test reactor

NASA Technical Reports Server (NTRS)

Putre, H. A.

1973-01-01

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

Westinghouse Small Modular Reactor nuclear steam supply system design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Memmott, M. J.; Harkness, A. W.; Van Wyk, J.

2012-07-01

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 first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development andmore » integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)« less

Technical Aspects Regarding the Management of Radioactive Waste from Decommissioning of Nuclear Facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dragolici, F.; Turcanu, C. N.; Rotarescu, G.

2003-02-25

The proper application of the nuclear techniques and technologies in Romania started in 1957, once with the commissioning of the Research Reactor VVR-S from IFIN-HH-Magurele. During the last 45 years, appear thousands of nuclear application units with extremely diverse profiles (research, biology, medicine, education, agriculture, transport, all types of industry) which used different nuclear facilities containing radioactive sources and generating a great variety of radioactive waste during the decommissioning after the operation lifetime is accomplished. A new aspect appears by the planning of VVR-S Research Reactor decommissioning which will be a new source of radioactive waste generated by decontamination, disassemblingmore » and demolition activities. By construction and exploitation of the Radioactive Waste Treatment Plant (STDR)--Magurele and the National Repository for Low and Intermediate Radioactive Waste (DNDR)--Baita, Bihor county, in Romania was solved the management of radioactive wastes arising from operation and decommissioning of small nuclear facilities, being assured the protection of the people and environment. The present paper makes a review of the present technical status of the Romanian waste management facilities, especially raising on treatment capabilities of ''problem'' wastes such as Ra-266, Pu-238, Am-241 Co-60, Co-57, Sr-90, Cs-137 sealed sources from industrial, research and medical applications. Also, contain a preliminary estimation of quantities and types of wastes, which would result during the decommissioning project of the VVR-S Research Reactor from IFIN-HH giving attention to some special category of wastes like aluminum, graphite and equipment, components and structures that became radioactive through neutron activation. After analyzing the technical and scientific potential of STDR and DNDR to handle big amounts of wastes resulting from the decommissioning of VVR-S Research Reactor and small nuclear facilities, the necessity of up-gradation of these nuclear objectives before starting the decommissioning plan is revealed. A short presentation of the up-grading needs is also presented.« less

DEMONSTRATION BULLETIN: FLAME REACTOR - HORSEHEAD RESOURCE DEVELOPMENT COMPANY, INC.

EPA Science Inventory

The Horsehead Resource Development Company, Inc. (HRD) Flame Reactor is a patented and proven high temperature thermal process designed to safely treat industrial residues and wastes containing metals. During processing, the waste material is introduced into the hottest portio...

METHOD AND APPARATUS FOR REACTOR SAFETY CONTROL

DOEpatents

Huston, N.E.

1961-06-01

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

Nuclear component horizontal seismic restraint

DOEpatents

Snyder, Glenn J.

1988-01-01

A nuclear component horizontal seismic restraint. Small gaps limit horizontal displacement of components during a seismic occurrence and therefore reduce dynamic loadings on the free lower end. The reactor vessel and reactor guard vessel use thicker section roll-forged rings welded between the vessel straight shell sections and the bottom hemispherical head sections. The inside of the reactor guard vessel ring forging contains local vertical dovetail slots and upper ledge pockets to mount and retain field fitted and installed blocks. As an option, the horizontal displacement of the reactor vessel core support cone can be limited by including shop fitted/installed local blocks in opposing alignment with the reactor vessel forged ring. Beams embedded in the wall of the reactor building protrude into apertures in the thermal insulation shell adjacent the reactor guard vessel ring and have motion limit blocks attached thereto to provide to a predetermined clearance between the blocks and reactor guard vessel ring.

Method and apparatus for enhancing reactor air-cooling system performance

DOEpatents

Hunsbedt, Anstein

1996-01-01

An enhanced decay heat removal system for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer.

Packed fluidized bed blanket for fusion reactor

DOEpatents

Chi, John W. H.

1984-01-01

A packed fluidized bed blanket for a fusion reactor providing for efficient radiation absorption for energy recovery, efficient neutron absorption for nuclear transformations, ease of blanket removal, processing and replacement, and on-line fueling/refueling. The blanket of the reactor contains a bed of stationary particles during reactor operation, cooled by a radial flow of coolant. During fueling/refueling, an axial flow is introduced into the bed in stages at various axial locations to fluidize the bed. When desired, the fluidization flow can be used to remove particles from the blanket.

Reactivity control assembly for nuclear reactor. [LMFBR

DOEpatents

Bollinger, L.R.

1982-03-17

This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

Removal of the Plutonium Recycle Test Reactor - 13031

DOE Office of Scientific and Technical Information (OSTI.GOV)

Herzog, C. Brad; Guercia, Rudolph; LaCome, Matt

2013-07-01

The 309 Facility housed the Plutonium Recycle Test Reactor (PRTR), an operating test reactor in the 300 Area at Hanford, Washington. The reactor first went critical in 1960 and was originally used for experiments under the Hanford Site Plutonium Fuels Utilization Program. The facility was decontaminated and decommissioned in 1988-1989, and the facility was deactivated in 1994. The 309 facility was added to Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) response actions as established in an Interim Record of Decision (IROD) and Action Memorandum (AM). The IROD directs a remedial action for the 309 facility, associated waste sites, associatedmore » underground piping and contaminated soils resulting from past unplanned releases. The AM directs a removal action through physical demolition of the facility, including removal of the reactor. Both CERCLA actions are implemented in accordance with U.S. EPA approved Remedial Action Work Plan, and the Remedial Design Report / Remedial Action Report associated with the Hanford 300-FF-2 Operable Unit. The selected method for remedy was to conventionally demolish above grade structures including the easily distinguished containment vessel dome, remove the PRTR and a minimum of 300 mm (12 in) of shielding as a single 560 Ton unit, and conventionally demolish the below grade structure. Initial sample core drilling in the Bio-Shield for radiological surveys showed evidence that the Bio-Shield was of sound structure. Core drills for the separation process of the PRTR from the 309 structure began at the deck level and revealed substantial thermal degradation of at least the top 1.2 m (4LF) of Bio-Shield structure. The degraded structure combined with the original materials used in the Bio-Shield would not allow for a stable structure to be extracted. The water used in the core drilling process proved to erode the sand mixture of the Bio-Shield leaving the steel aggregate to act as ball bearings against the core drill bit. A redesign is being completed to extract the 309 PRTR and entire Bio-Shield structure together as one monolith weighing 1100 Ton by cutting structural concrete supports. In addition, the PRTR has hundreds of contaminated process tubes and pipes that have to be severed to allow for a uniformly flush fit with a lower lifting frame. Thirty-two 50 mm (2 in) core drills must be connected with thirty-two wire saw cuts to allow for lifting columns to be inserted. Then eight primary saw cuts must be completed to severe the PRTR from the 309 Facility. Once the weight of the PRTR is transferred to the lifting frame, then the PRTR may be lifted out of the facility. The critical lift will be executed using four 450 Ton strand jacks mounted on a 9 m (30 LF) tall mobile lifting frame that will allow the PRTR to be transported by eight 600 mm (24 in) Slide Shoes. The PRTR will then be placed on a twenty-four line, double wide, self powered Goldhofer for transfer to the onsite CERCLA Disposal Cell (ERDF Facility), approximately 33 km (20 miles) away. (authors)« less

Liquefaction of calcium-containing subbituminous coals and coals of lower rank

DOEpatents

Brunson, Roy J.

1979-01-01

An improved process for the treatment of a calcium-containing subbituminous coal and coals of lower rank to form insoluble, thermally stable calcium salts which remain within the solids portions of the residue on liquefaction of the coal, thereby suppressing the formation of scale, made up largely of calcium carbonate which normally forms within the coal liquefaction reactor (i.e., coal liquefaction zone), e.g., on reactor surfaces, lines, auxiliary equipment and the like. An oxide of sulfur, in liquid phase, is contacted with a coal feed sufficient to impregnate the pores of the coal. The impregnated coal, in particulate form, can thereafter be liquefied in a coal liquefaction reactor (reaction zone) at coal liquefaction conditions without significant formation of scale.

Multiphysics Modeling for Dimensional Analysis of a Self-Heated Molten Regolith Electrolysis Reactor for Oxygen and Metals Production on the Moon and Mars

NASA Technical Reports Server (NTRS)

Dominguez, Jesus; Sibille, Laurent

2010-01-01

The technology of direct electrolysis of molten lunar regolith to produce oxygen and molten metal alloys has progressed greatly in the last few years. The development of long-lasting inert anodes and cathode designs as well as techniques for the removal of molten products from the reactor has been demonstrated. The containment of chemically aggressive oxide and metal melts is very difficult at the operating temperatures ca. 1600 C. Containing the molten oxides in a regolith shell can solve this technical issue and can be achieved by designing a self-heating reactor in which the electrolytic currents generate enough Joule heat to create a molten bath.

METHOD AND APPARATUS FOR IMPROVING PERFORMANCE OF A FAST REACTOR

DOEpatents

Koch, L.J.

1959-01-20

A specific arrangement of the fertile material and fissionable material in the active portion of a fast reactor to achieve improvement in performance and to effectively lower the operating temperatures in the center of the reactor is described. According to this invention a group of fuel elements containing fissionable material are assembled to form a hollow fuel core. Elements containing a fertile material, such as depleted uranium, are inserted into the interior of the fuel core to form a central blanket. Additional elemenis of fertile material are arranged about the fuel core to form outer blankets which in tunn are surrounded by a reflector. This arrangement of fuel core and blankets results in substantial flattening of the flux pattern.

Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor

NASA Astrophysics Data System (ADS)

Bury, Tomasz

2011-12-01

Gaseous hydrogen may be generated in a nuclear reactor system as an effect of the core overheating. This creates a risk of its uncontrolled combustion which may have a destructive consequences, as it could be observed during the Fukushima nuclear power plant accident. Favorable conditions for hydrogen production occur during heavy loss-of-coolant accidents. The author used an own computer code, called HEPCAL, of the lumped parameter type to realize a set of simulations of a large scale loss-of-coolant accidents scenarios within containment of second generation pressurized water reactor. Some simulations resulted in high pressure peaks, seemed to be irrational. A more detailed analysis and comparison with Three Mile Island and Fukushima accidents consequences allowed for withdrawing interesting conclusions.

METHOD AND APPARATUS FOR CONTROLLING NEUTRON DENSITY

DOEpatents

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

1961-06-27

A neutronic reactor comprising a moderator containing uniformly sized and spaced channels and uniformly dimensioned fuel elements is patented. The fuel elements have a fissionable core and an aluminum jacket. The cores and the jackets of the fuel elements in the central channels of the reactor are respectively thinner and thicker than the cores and jackets of the fuel elements in the remainder of the reactor, producing a flattened flux.

Compact power reactor

DOEpatents

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

1978-01-01

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

Versatile Oxide Films Protect FeCrAl Alloys Under Normal Operation and Accident Conditions in Light Water Power Reactors

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

2018-02-01

The US has currently a fleet of 99 nuclear power light water reactors which generate approximately 20% of the electricity consumed in the country. Near 90% of the reactors are at least 30 years old. There are incentives to make the existing reactors safer by using accident tolerant fuels (ATF). Compared to the standard UO2-zirconium-based system, ATF need to tolerate loss of active cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. Ferritic iron-chromium-aluminum (FeCrAl) alloys have been identified as an alternative to replace current zirconium alloys. They contain Fe (base) + 10-22 Cr + 4-6 Al and may contain smaller amounts of other elements such as molybdenum and traces of others. FeCrAl alloys offer outstanding resistance to attack by superheated steam by developing an alumina oxide on the surface in case of a loss of coolant accident like at Fukushima. FeCrAl alloys also perform well under normal operation conditions both in boiling water reactors and pressurized water reactors because they are protected by a thin oxide rich in chromium. Under normal operation condition, the key element is Cr and under accident conditions it is Al.

Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rule, K.; Scott, J.; Larson, S.

1995-12-31

The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methodsmore » for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.« less

Efficient H2O2/CH3COOH oxidative desulfurization/denitrification of liquid fuels in sonochemical flow-reactors.

PubMed

Calcio Gaudino, Emanuela; Carnaroglio, Diego; Boffa, Luisa; Cravotto, Giancarlo; Moreira, Elizabeth M; Nunes, Matheus A G; Dressler, Valderi L; Flores, Erico M M

2014-01-01

The oxidative desulfurization/denitrification of liquid fuels has been widely investigated as an alternative or complement to common catalytic hydrorefining. In this process, all oxidation reactions occur in the heterogeneous phase (the oil and the polar phase containing the oxidant) and therefore the optimization of mass and heat transfer is of crucial importance to enhancing the oxidation rate. This goal can be achieved by performing the reaction in suitable ultrasound (US) reactors. In fact, flow and loop US reactors stand out above classic batch US reactors thanks to their greater efficiency and flexibility as well as lower energy consumption. This paper describes an efficient sonochemical oxidation with H2O2/CH3COOH at flow rates ranging from 60 to 800 ml/min of both a model compound, dibenzotiophene (DBT), and of a mild hydro-treated diesel feedstock. Four different commercially available US loop reactors (single and multi-probe) were tested, two of which were developed in the authors' laboratory. Full DBT oxidation and efficient diesel feedstock desulfurization/denitrification were observed after the separation of the polar oxidized S/N-containing compounds (S≤5 ppmw, N≤1 ppmw). Our studies confirm that high-throughput US applications benefit greatly from flow-reactors. Copyright © 2013 Elsevier B.V. All rights reserved.

Effects of Laves phase particles on recovery and recrystallization behaviors of Nb-containing FeCrAl alloys

DOE PAGES

Sun, Zhiqian; Edmondson, Philip D.; Yamamoto, Yukinori

2017-11-15

The microstructures and mechanical properties of deformed and annealed Nb-containing FeCrAl alloys were investigated. Fine dispersion of Fe 2Nb-type Laves phase particles was observed in the bcc-Fe matrix after applying a thermomechanical treatment, especially along grain/subgrain boundaries, which effectively stabilized the recovered and recrystallized microstructures compared with the Nb-free FeCrAl alloy. The stability of recovered areas increased with Nb content up to 1 wt%. The recrystallized grain structure in Nb-containing FeCrAl alloys consisted of elongated grains along the rolling direction with a weak texture when annealed below 1100 °C. An abnormal relationship between recrystallized grain size and annealing temperature wasmore » found. Microstructural inhomogeneity in the deformed and annealed states was explained based on the Taylor factor. Annealed Nb-containing FeCrAl alloys showed a good combination of strength and ductility, which is desirable for their application as fuel cladding in light-water reactors.« less

CO2 capture by means of an enzyme-based reactor

NASA Technical Reports Server (NTRS)

Cowan, R. M.; Ge, J-J; Qin, Y-J; McGregor, M. L.; Trachtenberg, M. C.

2003-01-01

We report a means for efficient and selective extraction of carbon dioxide (CO(2)) at low to medium concentration from mixed gas streams. CO(2) capture was accomplished by use of a novel enzyme-based, facilitated transport contained liquid membrane (EBCLM) reactor. The parametric studies we report explore both structural and operational parameters of this design. The structural parameters include carbonic anhydrase (CA) concentration, buffer concentration and pH, and liquid membrane thickness. The operational parameters are temperature, humidity of the inlet gas stream, and CO(2) concentration in the feed stream. The data show that this system effectively captures CO(2) over the range 400 ppm to at least 100,000 ppm, at or around ambient temperature and pressure. In a single pass across this homogeneous catalyst design, given a feed of 0.1% CO(2), the selectivity of CO(2) versus N(2) is 1,090 : 1 and CO(2) versus O(2) is 790 :1. CO(2) permeance is 4.71 x 10(-8) molm(-2) Pa(-1) sec(-1). The CLM design results in a system that is very stable even in the presence of dry feed and sweep gases.

CHIMNEY FOR BOILING WATER REACTOR

DOEpatents

Petrick, M.

1961-08-01

A boiling-water reactor is described which has vertical fuel-containing channels for forming steam from water. Risers above the channels increase the head of water radially outward, whereby water is moved upward through the channels with greater force. The risers are concentric and the radial width of the space between them is somewhat small. There is a relatively low rate of flow of water up through the radially outer fuel-containing channels, with which the space between the risers is in communication. (AE C)

FUEL ELEMENT CONSTRUCTION

DOEpatents

Simnad, M.T.

1961-08-15

A method of preventing diffusible and volatile fission products from diffusing through a fuel element container and contaminating reactor coolant is described. More specifically, relatively volatile and diffusible fission products either are adsorbed by or react with magnesium fluoride or difluoride to form stable, less volatile, less diffusible forms. The magnesium fluoride or difluoride is disposed anywhere inwardly from the outer surface of the fuel element container in order to be contacted by the fission products before they reach and contaminate the reactor coolant. (AEC)

Treatment of semivolatile compounds in high strength wastes using an anaerobic expanded-bed GAC reactor

EPA Science Inventory

The potential of the anaerobic, expanded bed granular activated carbon (GAC) reactor in treating a high strength waste containing RCRA semivolatile organic compounds (VOCs) was studied. Six semivolatiles, orthochlorophenol, nitrobenzene, naphthalene, para-nitrophenol, lindane, a...

10 CFR 110.41 - Executive Branch review.

Code of Federal Regulations, 2011 CFR

2011-01-01

.... (6) An export involving assistance to end uses related to isotope separation, chemical reprocessing, heavy water production, advanced reactors, or the fabrication of nuclear fuel containing plutonium... equipment to a foreign reactor. (8) An export involving radioactive waste. (9) An export to any country...

Batch Tests To Determine Activity Distribution and Kinetic Parameters for Acetate Utilization in Expanded-Bed Anaerobic Reactors

PubMed Central

Fox, Peter; Suidan, Makram T.

1990-01-01

Batch tests to measure maximum acetate utilization rates were used to determine the distribution of acetate utilizers in expanded-bed sand and expanded-bed granular activated carbon (GAC) reactors. The reactors were fed a mixture of acetate and 3-ethylphenol, and they contained the same predominant aceticlastic methanogen, Methanothrix sp. Batch tests were performed both on the entire reactor contents and with media removed from the reactors. Results indicated that activity was evenly distributed within the GAC reactors, whereas in the sand reactor a sludge blanket on top of the sand bed contained approximately 50% of the activity. The Monod half-velocity constant (Ks) for the acetate-utilizing methanogens in two expanded-bed GAC reactors was searched for by combining steady-state results with batch test data. All parameters necessary to develop a model with Monod kinetics were experimentally determined except for Ks. However, Ks was a function of the effluent 3-ethylphenol concentration, and batch test results demonstrated that maximum acetate utilization rates were not a function of the effluent 3-ethylphenol concentration. Addition of a competitive inhibition term into the Monod expression predicted the dependence of Ks on the effluent 3-ethylphenol concentration. A two-parameter search determined a Ks of 8.99 mg of acetate per liter and a Ki of 2.41 mg of 3-ethylphenol per liter. Model predictions were in agreement with experimental observations for all effluent 3-ethylphenol concentrations. Batch tests measured the activity for a specific substrate and determined the distribution of activity in the reactor. The use of steady-state data in conjunction with batch test results reduced the number of unknown kinetic parameters and thereby reduced the uncertainty in the results and the assumptions made. PMID:16348175

Batch tests to determine activity distribution and kinetic parameters for acetate utilization in expanded-bed anaerobic reactors.

PubMed

Fox, P; Suidan, M T

1990-04-01

Batch tests to measure maximum acetate utilization rates were used to determine the distribution of acetate utilizers in expanded-bed sand and expanded-bed granular activated carbon (GAC) reactors. The reactors were fed a mixture of acetate and 3-ethylphenol, and they contained the same predominant aceticlastic methanogen, Methanothrix sp. Batch tests were performed both on the entire reactor contents and with media removed from the reactors. Results indicated that activity was evenly distributed within the GAC reactors, whereas in the sand reactor a sludge blanket on top of the sand bed contained approximately 50% of the activity. The Monod half-velocity constant (K(s)) for the acetate-utilizing methanogens in two expanded-bed GAC reactors was searched for by combining steady-state results with batch test data. All parameters necessary to develop a model with Monod kinetics were experimentally determined except for K(s). However, K(s) was a function of the effluent 3-ethylphenol concentration, and batch test results demonstrated that maximum acetate utilization rates were not a function of the effluent 3-ethylphenol concentration. Addition of a competitive inhibition term into the Monod expression predicted the dependence of K(s) on the effluent 3-ethylphenol concentration. A two-parameter search determined a K(s) of 8.99 mg of acetate per liter and a K(i) of 2.41 mg of 3-ethylphenol per liter. Model predictions were in agreement with experimental observations for all effluent 3-ethylphenol concentrations. Batch tests measured the activity for a specific substrate and determined the distribution of activity in the reactor. The use of steady-state data in conjunction with batch test results reduced the number of unknown kinetic parameters and thereby reduced the uncertainty in the results and the assumptions made.

Fluid sampling system for a nuclear reactor

DOEpatents

Lau, Louis K.; Alper, Naum I.

1994-01-01

A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump.

Fluid sampling system for a nuclear reactor

DOEpatents

Lau, L.K.; Alper, N.I.

1994-11-22

A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump. 1 fig.

Radiation effects on polymers for coatings on copper canisters used for the containment of radioactive materials

NASA Astrophysics Data System (ADS)

Mortley, Aba; Bonin, H. W.; Bui, V. T.

2008-05-01

The present work proposes applying polyurethane coatings as an additional barrier in the design of Canadian nuclear waste disposal containers. The goal of the present research is to investigate the physico-mechanical integrity of a natural castor oil-based polyurethane (COPU) to be used as a coating material in pH-radiation-temperature environments. As the first part to these inquiries, the present paper investigates the effect of a mixed radiation field supplied by a SLOWPOKE-2 nuclear research reactor on COPUs that differ only by their isocyanate structure. FTIR, DSC, DMA, WAXS, and MALDI are used to characterize the changes that occur as a result of radiation and to relate these changes to polymer structure and composition. The COPUs used in the present work have demonstrated sustained physico-mechanical properties up to accumulated doses of 2.0 MGy and are therefore suitable for end-uses in radiation environments such as those expected in the deep geological repository.

Coupled field-structural analysis of HGTR fuel brick using ABAQUS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, S.; Jain, R.; Majumdar, S.

2012-07-01

High-temperature, gas-cooled reactors (HTGRs) are usually helium-gas cooled, with a graphite core that can operate at reactor outlet temperatures much higher than can conventional light water reactors. In HTGRs, graphite components moderate and reflect neutrons. During reactor operation, high temperature and high irradiation cause damage to the graphite crystal and grains and create other defects. This cumulative structural damage during the reactor lifetime leads to changes in graphite properties, which can alter the ability to support the designed loads. The aim of the present research is to develop a finite-element code using commercially available ABAQUS software for the structural integritymore » analysis of graphite core components under extreme temperature and irradiation conditions. In addition, the Reactor Geometry Generator tool-kit, developed at Argonne National Laboratory, is used to generate finite-element mesh for complex geometries such as fuel bricks with multiple pin holes and coolant flow channels. This paper presents the proposed concept and discusses results of stress analysis simulations of a fuel block with H-451 grade material properties. (authors)« less

Noble gas, iodine, and cesium transport in a postulated loss of decay heat removal accident at Browns Ferry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wichner, R.P.; Hodge, S.A.; Weber, C.F.

1984-08-01

This report presents an analysis of the movement of noble gas, iodine, and cesium fission products within the Mark-I containment BWR reactor system represented by Browns Ferry Unit 1 during a postulated accident sequence initiated by a loss of decay heat removal capability following a scram. The event analysis showed that this accident could be brought under control by various means, but the sequence with no operator action ultimately leads to containment (drywell) failure followed by loss of water from the reactor vessel, core degradation due to overheating, and reactor vessel failure with attendant movement of core debris onto themore » drywell floor. The analysis of fission product transport presented in this report is based on the no-operator-action sequence and provides an estimate of fission product inventories, as a function of time, within 14 control volumes outside the core, with the atmosphere considered as the final control volume in the transport sequence. As in the case of accident sequences previously studied, we find small barrier for noble gas ejection to air, these gases being effectively purged from the drywell and reactor building by steam and concrete degradation gases. However, significant decay of krypton isotopes occurs during the long delay times involved in this sequence. In contrast, large degrees of holdup for iodine and cesium are projected due to the chemical reactivity of these elements. Only about 2 x 10/sup -4/% of the initial iodine and cesium activity are predicted to be released to the atmosphere. Principal barriers for release are deposition on reactor vessel and containment walls. A significant amount of iodine is captured in the water pool formed in the reactor building basement after actuation of the fire protection system.« less

Performance Study of Chromium (VI) Removal in Presence of Phenol in a Continuous Packed Bed Reactor by Escherichia coli Isolated from East Calcutta Wetlands

PubMed Central

Chakraborty, Bhaswati; Indra, Suvendu; Hazra, Ditipriya; Betai, Rupal; Ray, Lalitagauri; Basu, Srabanti

2013-01-01

Organic pollutants, like phenol, along with heavy metals, like chromium, are present in various industrial effluents that pose serious health hazard to humans. The present study looked at removal of chromium (VI) in presence of phenol in a counter-current continuous packed bed reactor packed with E. coli cells immobilized on clay chips. The cells removed 85% of 500 mg/L of chromium (VI) from MS media containing glucose. Glucose was then replaced by 500 mg/L phenol. Temperature and pH of the MS media prior to addition of phenol were 30°C and 7, respectively. Hydraulic retention times of phenol- and chromium (VI)-containing synthetic media and air flow rates were varied to study the removal efficiency of the reactor system. Then temperature conditions of the reactor system were varied from 10°C to 50°C, the optimum being 30°C. The pH of the media was varied from pH 1 to pH 12, and the optimum pH was found to be 7. The maximum removal efficiency of 77.7% was achieved for synthetic media containing phenol and chromium (VI) in the continuous reactor system at optimized conditions, namely, hydraulic retention time at 4.44 hr, air flow rate at 2.5 lpm, temperature at 30°C, and pH at 7. PMID:24073400

Nitrogen removal from purified swine wastewater using biogas by semi-partitioned reactor.

PubMed

Waki, Miyoko; Yokoyama, Hiroshi; Ogino, Akifumi; Suzuki, Kazuyoshi; Tanaka, Yasuo

2008-09-01

Nitrate and ammonium removal from purified swine wastewater using biogas and air was investigated in continuous reactor operation. A novel type of reactor, a semi-partitioned reactor (SPR), which enables a biological reaction using methane and oxygen in the water phase and discharges these unused gases separately, was operated with a varying gas supply rate. Successful removal of NO(3)(-) and NH(4)(+) was observed when biogas and air of 1L/min was supplied to an SPR of 9L water phase with a NO(2,3)(-)-N and NH(4)(+)-N removal rate of 0.10 g/L/day and 0.060 g/L/day, respectively. The original biogas contained an average of 77.2% methane, and the discharged biogas from the SPR contained an average of 76.9% of unused methane that was useable for energy like heat or electricity production. Methane was contained in the discharged air from the SPR at an average of 2.1%. When gas supply rates were raised to 2L/min and the nitrogen load was increased, NO(3)(-) concentration was decreased, but NO(2)(-) accumulated in the reactor and the NO(2,3)(-)-N and NH(4)(+)-N removal activity declined. To recover the activity, lowering of the nitrogen load and the gas supply rate was needed. This study shows that the SPR enables nitrogen removal from purified swine wastewater using biogas under limited gas supply condition.

Cluster structure of anaerobic aggregates of an expanded granular sludge bed reactor.

PubMed

Gonzalez-Gil, G; Lens, P N; Van Aelst, A; Van As, H; Versprille, A I; Lettinga, G

2001-08-01

The metabolic properties and ultrastructure of mesophilic aggregates from a full-scale expanded granular sludge bed reactor treating brewery wastewater are described. The aggregates had a very high methanogenic activity on acetate (17.19 mmol of CH(4)/g of volatile suspended solids [VSS].day or 1.1 g of CH(4) chemical oxygen demand/g of VSS.day). Fluorescent in situ hybridization using 16S rRNA probes of crushed granules showed that 70 and 30% of the cells belonged to the archaebacterial and eubacterial domains, respectively. The spherical aggregates were black but contained numerous whitish spots on their surfaces. Cross-sectioning these aggregates revealed that the white spots appeared to be white clusters embedded in a black matrix. The white clusters were found to develop simultaneously with the increase in diameter. Energy-dispersed X-ray analysis and back-scattered electron microscopy showed that the whitish clusters contained mainly organic matter and no inorganic calcium precipitates. The white clusters had a higher density than the black matrix, as evidenced by the denser cell arrangement observed by high-magnification electron microscopy and the significantly higher effective diffusion coefficient determined by nuclear magnetic resonance imaging. High-magnification electron microscopy indicated a segregation of acetate-utilizing methanogens (Methanosaeta spp.) in the white clusters from syntrophic species and hydrogenotrophic methanogens (Methanobacterium-like and Methanospirillum-like organisms) in the black matrix. A number of physical and microbial ecology reasons for the observed structure are proposed, including the advantage of segregation for high-rate degradation of syntrophic substrates.

Cluster Structure of Anaerobic Aggregates of an Expanded Granular Sludge Bed Reactor

PubMed Central

Gonzalez-Gil, G.; Lens, P. N. L.; Van Aelst, A.; Van As, H.; Versprille, A. I.; Lettinga, G.

2001-01-01

The metabolic properties and ultrastructure of mesophilic aggregates from a full-scale expanded granular sludge bed reactor treating brewery wastewater are described. The aggregates had a very high methanogenic activity on acetate (17.19 mmol of CH4/g of volatile suspended solids [VSS]·day or 1.1 g of CH4 chemical oxygen demand/g of VSS·day). Fluorescent in situ hybridization using 16S rRNA probes of crushed granules showed that 70 and 30% of the cells belonged to the archaebacterial and eubacterial domains, respectively. The spherical aggregates were black but contained numerous whitish spots on their surfaces. Cross-sectioning these aggregates revealed that the white spots appeared to be white clusters embedded in a black matrix. The white clusters were found to develop simultaneously with the increase in diameter. Energy-dispersed X-ray analysis and back-scattered electron microscopy showed that the whitish clusters contained mainly organic matter and no inorganic calcium precipitates. The white clusters had a higher density than the black matrix, as evidenced by the denser cell arrangement observed by high-magnification electron microscopy and the significantly higher effective diffusion coefficient determined by nuclear magnetic resonance imaging. High-magnification electron microscopy indicated a segregation of acetate-utilizing methanogens (Methanosaeta spp.) in the white clusters from syntrophic species and hydrogenotrophic methanogens (Methanobacterium-like and Methanospirillum-like organisms) in the black matrix. A number of physical and microbial ecology reasons for the observed structure are proposed, including the advantage of segregation for high-rate degradation of syntrophic substrates. PMID:11472948

NUCLEAR REACTOR CORE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Preece, G.E.; Bell, F.R.; Page, R.W.

1963-03-01

A nuclear reactor core is described. It contains fuel in the form of blocks or pellets that have a grooved, wrinkled, or corrugated surface to provide a greater radiating surface area. The surfaces of spaces in the core are correspondingly corrugated for maximum heat exchange area. (C.E.S.)

TREATMENT OF VOCS IN HIGH STRENGTH WASTES USING AN ANAEROBIC EXPANDED-BED GAS REACTOR

EPA Science Inventory

The potential of the expanded-bed granular activated carbon (GAC) anaerobic reactor in treating a high strength waste containing RCRA volatile organic compounds (VOCs) was studied. A total of six VOCs, methylene chloride, chlorobenzene, carbon tetrachloride, chloroform, toluene ...

ANAEROBIC/AEROBIC BIODEGRADATION OF PENTACHLOROPHENOL USING GAC FLUIDIXED BED REACTORS: OPTIMIZATION OF THE EMPTY BED CONTACT TIME

EPA Science Inventory

An integrated reactor system has been developed to remediate pentachlorophenol (PCP) containing wastes using sequential anaerobic and aerobic biodegradation. Anaerobically, PCP was degraded to approximately equimolar concentrations (>99%) of chlorophenol (CP) in a granular activa...

A disposable, self-contained PCR chip.

PubMed

Kim, Jitae; Byun, Doyoung; Mauk, Michael G; Bau, Haim H

2009-02-21

A disposable, self-contained polymerase chain reaction (PCR) chip with on-board stored, just-on-time releasable, paraffin-passivated, dry reagents is described. During both storage and sample preparation, the paraffin immobilizes and protects the stored reagents. Fluid flow through the reactor leaves the reagents undisturbed. Prior to the amplification step, the chamber is filled with target analyte suspended in water. Upon heating the PCR chamber to the DNA's denaturation temperature, the paraffin melts and moves out of the way, and the reagents are released and hydrated. To better understand the reagent release process, a scaled up model of the reactor was constructed and the paraffin migration was visualized. Experiments were carried out with a 30 microl reactor demonstrating detectable amplification (with agarose gel electrophoresis) of 10 fg ( approximately 200 copies) of lambda DNA template. The in-reactor storage and on-time release of the PCR reagents reduce the number of needed operations and significantly simplifies the flow control that would, otherwise, be needed in lab-on-chip devices.

A Disposable, Self-Contained PCR Chip

PubMed Central

Kim, Jitae; Byun, Doyoung; Mauk, Michael G.; Bau, Haim H.

2009-01-01

A disposable, self-contained polymerase chain reaction (PCR) chip with on-board stored, just on time releasable, paraffin-passivated, dry reagents is described. During both storage and sample preparation, the paraffin immobilizes and protects the stored reagents. Fluid flow through the reactor leaves the reagents undisturbed. Prior to the amplification step, the chamber is filled with target analyte suspended in water. Upon heating the PCR chamber to the DNA’s denaturation temperature, the paraffin melts and moves out of the way, and the reagents are released and hydrated. To better understand the reagent release process, a scaled up model of the reactor was constructed and the paraffin migration was visualized. Experiments were carried out with a 30 μl reactor demonstrating detectable amplification (with agarose gel electrophoresis) of 10 fg (~200 copies) of lambda DNA template. The in-reactor storage and on-time release of the PCR reagents reduce the number of needed operations and significantly simplify the flow control that would, otherwise, be needed in lab-on-chip devices. PMID:19190797

Multi-Megawatt Space Nuclear Power Generation

DTIC Science & Technology

1993-06-28

electric generation, both for open- and closed-cycle opera- tion. These reactors use the particulate fuel of the type developed for HTGR reactors. What...commercial HTGR power reactors, the particles are held in place and directly cooled. Figure 2.7 shows the two types of fuel particles developed for...of MW(e), for pulsed energy devices. The FBR would use HTGR -type particle fuel , contained in a annular bed be- tween two porous frits. Helium would

Reactor for fluidized bed silane decomposition

NASA Technical Reports Server (NTRS)

Iya, Sridhar K. (Inventor)

1989-01-01

An improved heated fluidized bed reactor and method for the production of high purity polycrystalline silicon by silane pyrolysis wherein silicon seed particles are heated in an upper heating zone of the reactor and admixed with particles in a lower zone, in which zone a silane-containing gas stream, having passed through a lower cooled gas distribution zone not conducive to silane pyrolysis, contacts the heated seed particles whereon the silane is heterogeneously reduced to silicon.

Process for treating effluent from a supercritical water oxidation reactor

DOEpatents

Barnes, Charles M.; Shapiro, Carolyn

1997-01-01

A method for treating a gaseous effluent from a supercritical water oxidation reactor containing entrained solids is provided comprising the steps of expanding the gas/solids effluent from a first to a second lower pressure at a temperature at which no liquid condenses; separating the solids from the gas effluent; neutralizing the effluent to remove any acid gases; condensing the effluent; and retaining the purified effluent to the supercritical water oxidation reactor.

MOLTEN FLUORIDE NUCLEAR REACTOR FUEL

DOEpatents

Barton, C.J.; Grimes, W.R.

1960-01-01

Molten-salt reactor fuel compositions consisting of mixtures of fluoride salts are reported. In its broadest form, the composition contains an alkali fluoride such as sodium fluoride, zirconium tetrafluoride, and a uranium fluoride, the latter being the tetrafluoride or trifluoride or a mixture of the two. An outstanding property of these fuel compositions is a high coeffieient of thermal expansion which provides a negative temperature coefficient of reactivity in reactors in which they are used.

NEUTRONIC REACTORS

DOEpatents

Vernon, H.C.

1959-01-13

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

Tritium well depth, tritium well time and sponge mechanism for reducing tritium retention

NASA Astrophysics Data System (ADS)

Deng, B. Q.; Li, Z. X.; Li, C. Y.; Feng, K. M.

2011-07-01

New simulation results are predicted in a fusion reactor operation process. They are somewhat similar to, but quite different from, the xenon poisoning effects resulting from fission-produced iodine during the restart-up process of a fission reactor. We obtained completely new results of tritium well depth and tritium well time in magnetic confinement fusion energy research area. This study is carried out to investigate the following: what will be the least amount of tritium storage required to start up a fusion reactor and how long the fusion reactor needs to be operated for achieving the tritium break-even during the initial start-up phase due to the finite tritium-breeding time, which is dependent on the tritium breeder, specific structure of the breeding zone, layout of the coolant flow pipes, tritium recovery scheme and applied extraction process, the tritium retention of plasma facing component (PFC) and other reactor components, unrecoverable tritium fraction in the breeder, leakage to the inertial gas container and the natural radioactive decay time constant. We describe these new issues and answer these problems by setting up and solving a set of equations, which are described by a dynamic subsystem model of tritium inventory evolution in a fusion experimental breeder (FEB). Reasonable results are obtained using our simulation model. It is found that the tritium well depth is about 0.319 kg and the tritium well time is approximately 235 full power operation days for the reference case of the designed FEB configuration, and it is also found that after one-year operation the tritium storage reaches 0.767 kg, which is more than the least amount of tritium storage required to start up another FEB-like fusion reactor. The results show that the tritium retention in the PFC is equivalent to 11.9% of tritium well depth that is fairly consistent with the result of 10-20% deduced from the integrated particle balance of European tokamaks. Based on our experimental and theoretical studies, some new mechanisms are proposed for reducing the tritium retention in PFC and structure materials of tritium-breeding blanket. In this paper, a qualitative analysis of the 'sponge effect' is carried out. The 'sponge effect' may help us to reduce tritium retention by ~20% in the PFC.

ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

E. Blanford; E. Keldrauk; M. Laufer

2010-09-20

Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement,more » and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using factory prefabricated structural modules, for application to external event shell and base isolated structures.« less

REACTOR FUEL SCAVENGING MEANS

DOEpatents

Coffinberry, A.S.

1962-04-10

A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

Fuel leak detection apparatus for gas cooled nuclear reactors

DOEpatents

Burnette, Richard D.

1977-01-01

Apparatus is disclosed for detecting nuclear fuel leaks within nuclear power system reactors, such as high temperature gas cooled reactors. The apparatus includes a probe assembly that is inserted into the high temperature reactor coolant gaseous stream. The probe has an aperture adapted to communicate gaseous fluid between its inside and outside surfaces and also contains an inner tube for sampling gaseous fluid present near the aperture. A high pressure supply of noncontaminated gas is provided to selectively balance the pressure of the stream being sampled to prevent gas from entering the probe through the aperture. The apparatus includes valves that are operable to cause various directional flows and pressures, which valves are located outside of the reactor walls to permit maintenance work and the like to be performed without shutting down the reactor.

Sludge granulation and performance of a low superficial gas velocity sequencing batch reactor (SBR) in the treatment of prepared sanitary wastewater.

PubMed

Ji, Guodong; Zhai, Fengmin; Wang, Rongjing; Ni, Jinren

2010-12-01

A sequencing batch reactor (SBR) employing a low superficial gas velocity was used to produce aerobic granular sludge for wastewater treatment. At a gas velocity of 0.0056 m s(-1) sludge containing a mixture of light yellow and black granules with similar functional groups was quickly formed. The black granules contained crystals of CaCO(3), FeS, and Fe(2)O(3) as well as filamentous bacteria that strengthened the particles and reduced the mass transfer resistance. No inorganic crystals were detected in the yellow sludge granules, and their structure was maintained through cohesion mediated by extracellular polymeric substances (EPS). The light yellow granules were denser and offered better settling performance than the black granules, enhancing the settling properties of the mixed sludge. During a 12-h cycle, the maximum reductions in chemical oxygen demand (COD), NH(3)-N, and total nitrogen (TN) occurred at 240, 480, and 360 min with removal efficiencies of 90%, 90%, and 54%. When the cycle time was limited to 480 min, self-dissolution of the granules was avoided while sill maintaining removal efficiencies for COD, NH(3)-N, and TN of 88%, 90%, and 53%. 2010 Elsevier Ltd. All rights reserved.

The aftermath of the Fukushima nuclear accident: Measures to contain groundwater contamination.

PubMed

Gallardo, Adrian H; Marui, Atsunao

2016-03-15

Several measures are being implemented to control groundwater contamination at the Fukushima Daiichi Nuclear Plant. This paper presents an overview of work undertaken to contain the spread of radionuclides, and to mitigate releases to the ocean via hydrological pathways. As a first response, contaminated water is being held in tanks while awaiting treatment. Limited storage capacity and the risk of leakage make the measure unsustainable in the long term. Thus, an impervious barrier has been combined with a drain system to minimize the discharge of groundwater offshore. Caesium in seawater at the plant port has largely dropped, although some elevated concentrations are occasionally recorded. Moreover, a dissimilar decline of the radioactivity in fish could indicate additional sources of radionuclides intake. An underground frozen shield is also being constructed around the reactors. This structure would reduce inflows to the reactors and limit the interaction between fresh and contaminated waters. Additional strategies include groundwater abstraction and paving of surfaces to lower water levels and further restrict the mobilisation of radionuclides. Technical difficulties and public distrust pose an unprecedented challenge to the site remediation. Nevertheless, the knowledge acquired during the initial work offers opportunities for better planning and more rigorous decisions in the future. Copyright © 2016 Elsevier B.V. All rights reserved.

MYRRHA: A multipurpose nuclear research facility

NASA Astrophysics Data System (ADS)

Baeten, P.; Schyns, M.; Fernandez, Rafaël; De Bruyn, Didier; Van den Eynde, Gert

2014-12-01

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a multipurpose research facility currently being developed at SCK•CEN. MYRRHA is based on the ADS (Accelerator Driven System) concept where a proton accelerator, a spallation target and a subcritical reactor are coupled. MYRRHA will demonstrate the ADS full concept by coupling these three components at a reasonable power level to allow operation feedback. As a flexible irradiation facility, the MYRRHA research facility will be able to work in both critical as subcritical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV and fusion reactors, and radioisotope production for medical and industrial applications. MYRRHA will be cooled by lead-bismuth eutectic and will play an important role in the development of the Pb-alloys technology needed for the LFR (Lead Fast Reactor) GEN IV concept. MYRRHA will also contribute to the study of partitioning and transmutation of high-level waste. Transmutation of minor actinides (MA) can be completed in an efficient way in fast neutron spectrum facilities, so both critical reactors and subcritical ADS are potential candidates as dedicated transmutation systems. However critical reactors heavily loaded with fuel containing large amounts of MA pose reactivity control problems, and thus safety problems. A subcritical ADS operates in a flexible and safe manner, even with a core loading containing a high amount of MA leading to a high transmutation rate. In this paper, the most recent developments in the design of the MYRRHA facility are presented.

Bioaugmentation of activated sludge towards 3-chloroaniline removal with a mixed bacterial population carrying a degradative plasmid.

PubMed

Bathe, Stephan; Schwarzenbeck, Norbert; Hausner, Martina

2009-06-01

A bioaugmentation approach combining several strategies was applied to achieve degradation of 3-chloroaniline (3CA) in semicontinuous activated sludge reactors. In a first step, a 3CA-degrading Comamonas testosteroni strain carrying the degradative plasmid pNB2 was added to a biofilm reactor, and complete 3CA degradation together with spread of the plasmid within the indigenous biofilm population was achieved. A second set of reactors was then bioaugmented with either a suspension of biofilm cells removed from the carrier material or with biofilm-containing carrier material. 3CA degradation was established rapidly in all bioaugmented reactors, followed by a slow adaptation of the non-bioaugmented control reactors. In response to variations in 3CA concentration, all reactors exhibited temporary performance breakdowns. Whereas duplicates of the control reactors deviated in their behaviour, the bioaugmented reactors appeared more reproducible in their performance and population dynamics. Finally, the carrier-bioaugmented reactors showed an improved performance in the presence of high 3CA influent concentrations over the suspension-bioaugmented reactors. In contrast, degradation in one control reactor failed completely, but was rapidly established in the remaining control reactor.

Characterization of an immobilized cell, trickle bed reactor during long term butanol (ABE) fermentation.

PubMed

Park, C H; Okos, M R; Wankat, P C

1990-06-20

Acetone-butanol-ethanol (ABE) fermentation was performed continuously in an immobilized cell, trickle bed reactor for 54 days without, degeneration by maintaining the pH above 4.3. Column clogging was minimized by structured packing of immobilization matrix. The reactor contained two serial glass columns packed with Clostridium acetobutylicum adsorbed on 12- and 20-in.-long polyester sponge strips at total flow rates between 38 and 98.7 mL/h. Cells were initially grown at 20 g/L glucose resulting in low butanol (1.15 g/L) production encouraging cell growth. After the initial cell growth phase a higher glucose concentration (38.7 g/L) improved solvent yield from 13.2 to 24.1 wt%, and butanol production rate was the best. Further improvement in solvent yield and butanol production rate was not observed with 60 g/L of glucose. However, when the fresh nutrient supply was limited to only the first column, solvent yield increased to 27.3 wt% and butanol selectivity was improved to 0.592 as compared to 0.541 when fresh feed was fed to both columns. The highest butanol concentration of 5.2 g/L occurred at 55% conversion of the feed with 60 g/L glucose. Liquid product yield of immobilized cells approached the theoretical value reported in the literature. Glucose and product concentration profiles along the column showed that the columns can be divided into production and inhibition regions. The length of each zone was dependent upon the feed glucose concentration and feed pattern. Unlike batch fermentation, there was no clear distinction between acid and solvent production regions. The pH dropped, from 6.18-6.43 to 4.50-4.90 in the first inch of the reactor. The pH dropped further to 4.36-4.65 by the exit of the column. The results indicate that the strategy for long term stable operation with high solvent yield requires a structured packing of biologically stable porous matrix such as polyester sponge, a pH maintenance above 4.3, glucose concentrations up to 60 g/L and nutrient supply only to the inlet of the reactor.

Modular fabrication and characterization of complex silicon carbide composite structures Advanced Reactor Technologies (ART) Research Final Report (Feb 2015 – May 2017)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Khalifa, Hesham

Advanced ceramic materials exhibit properties that enable safety and fuel cycle efficiency improvements in advanced nuclear reactors. In order to fully exploit these desirable properties, new processing techniques are required to produce the complex geometries inherent to nuclear fuel assemblies and support structures. Through this project, the state of complex SiC-SiC composite fabrication for nuclear components has advanced significantly. New methods to produce complex SiC-SiC composite structures have been demonstrated in the form factors needed for in-core structural components in advanced high temperature nuclear reactors. Advanced characterization techniques have been employed to demonstrate that these complex SiC-SiC composite structures providemore » the strength, toughness and hermeticity required for service in harsh reactor conditions. The complex structures produced in this project represent a significant step forward in leveraging the excellent high temperature strength, resistance to neutron induced damage, and low neutron cross section of silicon carbide in nuclear applications.« less

Development work for a borax internal core-catcher for a gas-cooled fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Donne, M.D.; Dorner, S.; Schumacher, G.

1978-07-01

Preliminary thermal calculations show that a corecatcher, which is able to cope with the complete meltdown of the core and blankets of a 1000-MW(electric) gas-cooled fast reactor, appears to be feasible. This core-catcher is based on borax (Na/sub 2/B/sub 4/O/sub 7/) dissolving the oxide fuel and the fission products occurring in oxide form. The borax is contained in steel boxes forming a 2.2-m-thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel (PCRV), just underneath the reactor core. After a complete meltdown accident, the fission products, in oxide form, are dispersed in the pool formedmore » by the liquid borax. The metallic fission products are contained in the steel lying below the borax pool and in contact with the water-cooled PCRV liner. The volumetric power density of the molten core is conveniently reduced as it is dissolved in the borax, and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system.« less

Pressurized fluidized bed reactor

DOEpatents

Isaksson, J.

1996-03-19

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

Pressurized fluidized bed reactor

DOEpatents

Isaksson, Juhani

1996-01-01

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

The Development of Neutron Radiography and Tomography on a SLOWPOKE-2 Reactor

NASA Astrophysics Data System (ADS)

Bennett, L. G. I.; Lewis, W. J.; Hungler, P. C.

Development of neutron radiography at the Royal Military College of Canada (RMC) started by trying to interest the Royal Canadian Air Force (RCAF) in this new non-destructive testing (NDT) technique. A Californium-252 based device was ordered and then installed at RMC for development of applicable techniques for aircraft by the first author. A second and transportable device was then designed, modified and used in trials at RCAF Bases and other locations for one year. This activity was the only foreign loan of the U.S. Californium Loan Program. Around this time, SLOWPOKE-2 reactors were being installed at four Canadian universities, while a new science and engineering building was being built at RMC. A reactor pool was incorporated and efforts to procure a reactor succeeded a decade later with a SLOWPOKE-2 reactor being installed at RMC. The only modification by the vendor for RMC was a thermal column replacing an irradiation site inside the reactor container for a later installation of a neutron beam tube (NBT). Development of a working NBT took several years, starting with the second author. A demonstration of the actual worth of neutron radiography took place with a CF-18 Hornet aircraft being neutron and X-radiographed at McClellan Air Force Base, Sacramento, CA. This inspection was followed by one of the rudders that had indications of water ingress being radiographed successfully at RMC just after the NBT became functional. The next step was to develop a neutron radioscopy system (NRS), initially employing film and then digital imaging, and is in use today for all flight control surfaces (FCS). With the third author, a technique capable of removing water from affected FCS was developed at RMC. Heating equipment and a vacuum system were utilized to carefully remove the water. This technique was proven using a sequence of near real time neutron images obtained during the drying process. The results of the drying process were correlated with a relative humidity gauge and an NDT technique that could be performed at Canadian Forces (CF) Bases was developed. In order to determine the structural integrity of the component having undergone this water removal, further research was required into the effect of water inside composite honeycomb structures. This need has led to the present development of neutron tomography on the reactor at RMC, which is capable of determining the exact location of water ingress inside composite components. This technique has been successfully applied to coupons as well as to complete rudders.

Molten uranium dioxide structure and dynamics

DOE PAGES

Skinner, L. B.; Parise, J. B.; Benmore, C. J.; ...

2014-11-21

Uranium dioxide (UO 2) is the major nuclear fuel component of fission power reactors. A key concern during severe accidents is the melting and leakage of radioactive UO 2 as it corrodes through its zirconium cladding and steel containment. Yet, the very high temperatures (>3140 kelvin) and chemical reactivity of molten UO 2 have prevented structural studies. In this work, we combine laser heating, sample levitation, and synchrotron x-rays to obtain pair distribution function measurements of hot solid and molten UO 2. The hot solid shows a substantial increase in oxygen disorder around the lambda transition (2670 K) but negligiblemore » U-O coordination change. On melting, the average U-O coordination drops from 8 to 6.7 ± 0.5. Molecular dynamics models refined to this structure predict higher U-U mobility than 8-coordinated melts.« less

NEUTRONIC REACTOR CORE INSTRUMENT

DOEpatents

Mims, L.S.

1961-08-22

A multi-purpose instrument for measuring neutron flux, coolant flow rate, and coolant temperature in a nuclear reactor is described. The device consists essentially of a hollow thimble containing a heat conducting element protruding from the inner wall, the element containing on its innermost end an amount of fissionsble materinl to function as a heat source when subjected to neutron flux irradiation. Thermocouple type temperature sensing means are placed on the heat conducting element adjacent the fissionable material and at a point spaced therefrom, and at a point on the thimble which is in contact with the coolant fluid. The temperature differentials measured between the thermocouples are determinative of the neutron flux, coolant flow, and temperature being measured. The device may be utilized as a probe or may be incorporated in a reactor core. (AE C)

The Liquid Annular Reactor System (LARS) propulsion

NASA Technical Reports Server (NTRS)

Powell, James; Ludewig, Hans; Horn, Frederick; Lenard, Roger

1990-01-01

A concept for very high specific impulse (greater than 2000 seconds) direct nuclear propulsion is described. The concept, termed the liquid annular reactor system (LARS), uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (approximately 6000 K). Operating pressure is moderate (approximately 10 atm), with the result that the outlet hydrogen is virtually 100 percent dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use seven rotating fuel elements, are beryllium moderated, and have critical radii of approximately 100 cm (core L/D approximately equal to 1.5).

Design of a new reactor-like high temperature near ambient pressure scanning tunneling microscope for catalysis studies.

PubMed

Tao, Franklin Feng; Nguyen, Luan; Zhang, Shiran

2013-03-01

Here, we present the design of a new reactor-like high-temperature near ambient pressure scanning tunneling microscope (HT-NAP-STM) for catalysis studies. This HT-NAP-STM was designed for exploration of structures of catalyst surfaces at atomic scale during catalysis or under reaction conditions. In this HT-NAP-STM, the minimized reactor with a volume of reactant gases of ∼10 ml is thermally isolated from the STM room through a shielding dome installed between the reactor and STM room. An aperture on the dome was made to allow tip to approach to or retract from a catalyst surface in the reactor. This dome minimizes thermal diffusion from hot gas of the reactor to the STM room and thus remains STM head at a constant temperature near to room temperature, allowing observation of surface structures at atomic scale under reaction conditions or during catalysis with minimized thermal drift. The integrated quadrupole mass spectrometer can simultaneously measure products during visualization of surface structure of a catalyst. This synergy allows building an intrinsic correlation between surface structure and its catalytic performance. This correlation offers important insights for understanding of catalysis. Tests were done on graphite in ambient environment, Pt(111) in CO, graphene on Ru(0001) in UHV at high temperature and gaseous environment at high temperature. Atom-resolved surface structure of graphene on Ru(0001) at 500 K in a gaseous environment of 25 Torr was identified.

Nuclear reactors built, being built, or planned 1993

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1993-08-01

Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1993. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical datamore » that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: civilian, production, military, export, and critical assembly.« less

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

NASA Astrophysics Data System (ADS)

Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

2006-01-01

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.

Nuclear reactor reflector

DOEpatents

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

1994-01-01

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

Nuclear reactor reflector

DOEpatents

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

1994-06-07

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

Method and apparatus for enhancing reactor air-cooling system performance

DOEpatents

Hunsbedt, A.

1996-03-12

An enhanced decay heat removal system is disclosed for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer. 6 figs.

NEUTRON REACTOR HAVING A Xe$sup 135$ SHIELD

DOEpatents

Stanton, H.E.

1957-10-29

Shielding for reactors of the type in which the fuel is a chain reacting liquid composition comprised essentially of a slurry of fissionable and fertile material suspended in a liquid moderator is discussed. The neutron reflector comprises a tank containing heavy water surrounding the reactor, a shield tank surrounding the reflector, a gamma ray shield surrounding said shield tank, and a means for conveying gaseous fission products, particularly Xe/sup 135/, from the reactor chamber to the shield tank, thereby serving as a neutron shield by capturing the thermalized neutrons that leak outwardly from the shield tank.

27. The top of a typical pile, F Reactor in ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

27. The top of a typical pile, F Reactor in February 1945 in this case, showing the vertical safety rods (VSRs) and the cables that support them. The rods could be dropped into the pile to effect a rapid shutdown. The four silvered-colored drums on the left contained boron solution and are part of the last ditch safety system. Should the VSRs channels become blocked by an occurrence such as an earthquake, the solution could be dumped into the VSR channels to help shut down the reactor. D-8334 - B Reactor, Richland, Benton County, WA

IMPROVEMENTS RELATING TO NUCLEAR REACTOR CORE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bell, F.R.

1963-03-01

A nuclear reactor core composed of a number of stacked horizontal layers is described. Each layer is made up of elements of moderator material of equal height and of generally hexagonal cross-section. Each element has holes containing nuclear fuel and separate ones for coolant. (C.E.S.)

10 CFR 110.41 - Executive Branch review.

Code of Federal Regulations, 2010 CFR

2010-01-01

... export involving assistance to end uses related to isotope separation, chemical reprocessing, heavy water production, advanced reactors, or the fabrication of nuclear fuel containing plutonium, except for exports of... foreign reactor. (8) An export involving radioactive waste. (9) An export to any country listed in § 110...

NON-CORROSIVE REACTOR FUEL SYSTEM

DOEpatents

Herrick, C.C.

1962-08-14

A non-corrosive nuclear reactor fuel system was developed utilizing a molten plutonium-- iron alloy fuel having about 2 at.% carbon and contained in a tantalum vessel. This carbon reacts with the interior surface of the tantalum vessel to form a plutonium resistant self-healing tantalum carbide film. (AEC)

Hydrolysis and fractionation of lignocellulosic biomass

DOEpatents

Torget, Robert W.; Padukone, Nandan; Hatzis, Christos; Wyman, Charles E.

2000-01-01

A multi-function process is described for the hydrolysis and fractionation of lignocellulosic biomass to separate hemicellulosic sugars from other biomass components such as extractives and proteins; a portion of the solubilized lignin; cellulose; glucose derived from cellulose; and insoluble lignin from said biomass comprising one or more of the following: optionally, as function 1, introducing a dilute acid of pH 1.0-5.0 into a continual shrinking bed reactor containing a lignocellulosic biomass material at a temperature of about 94 to about 160.degree. C. for a period of about 10 to about 120 minutes at a volumetric flow rate of about 1 to about 5 reactor volumes to effect solubilization of extractives, lignin, and protein by keeping the solid to liquid ratio constant throughout the solubilization process; as function 2, introducing a dilute acid of pH 1.0-5.0, either as virgin acid or an acidic stream from another function, into a continual shrinking bed reactor containing either fresh biomass or the partially fractionated lignocellulosic biomass material from function 1 at a temperature of about 94-220.degree. C. for a period of about 10 to about 60 minutes at a volumetric flow rate of about 1 to about 5 reactor volumes to effect solubilization of hemicellulosic sugars, semisoluble sugars and other compounds, and amorphous glucans by keeping the solid to liquid ratio constant throughout the solubilization process; as function 3, optionally, introducing a dilute acid of pH 1.0-5.0 either as virgin acid or an acidic stream from another function, into a continual shrinking bed reactor containing the partially fractionated lignocellulosic biomass material from function 2 at a temperature of about 180-280.degree. C. for a period of about 10 to about 60 minutes at a volumetric flow rate of 1 to about 5 reactor volumes to effect solubilization of cellulosic sugars by keeping the solid to liquid ratio constant throughout the solubilization process; and as function 4, optionally, introducing a dilute acid of pH 1.0-5.0 either as virgin acid or an acidic stream from another function, into a continual shrinking bed reactor containing the partially fractionated lignocellulosic biomass material from function 3 at a temperature of about 180-280.degree. C. for a period of about 10 to about 60 minutes at a volumetric flow rate of about 1 to about 5 reactor volumes to effect solubilization of cellulosic sugars by keeping the solid to liquid ratio constant throughout the solubilization process.

Irradiation Microstructure of Austenitic Steels and Cast Steels Irradiated in the BOR-60 Reactor at 320°C

NASA Astrophysics Data System (ADS)

Yang, Yong; Chen, Yiren; Huang, Yina; Allen, Todd; Rao, Appajosula

Reactor internal components are subjected to neutron irradiation in light water reactors, and with the aging of nuclear power plants around the world, irradiation-induced material degradations are of concern for reactor internals. Irradiation-induced defects resulting from displacement damage are critical for understanding degradation in structural materials. In the present work, microstructural changes due to irradiation in austenitic stainless steels and cast steels were characterized using transmission electron microscopy. The specimens were irradiated in the BOR-60 reactor, a fast breeder reactor, up to 40 dpa at 320°C. The dose rate was approximately 9.4x10-7 dpa/s. Void swelling and irradiation defects were analyzed for these specimens. A high density of faulted loops dominated the irradiated-altered microstructures. Along with previous TEM results, a dose dependence of the defect structure was established at 320°C.

Hanging core support system for a nuclear reactor. [LMFBR

DOEpatents

Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

1984-04-26

For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

DOE PAGES

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.; ...

2016-12-21

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

Design of a heatpipe-cooled Mars-surface fission reactor

NASA Astrophysics Data System (ADS)

Poston, David I.; Kapernick, Richard J.; Guffee, Ray M.; Reid, Robert S.; Lipinski, Ronald J.; Wright, Steven A.; Talandis, Regina A.

2002-01-01

The next generation of robotic missions to Mars will most likely require robust power sources in the range of 3 to 20 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The goal of this study is to design a compact, low-mass fission system that meets Mars-surface power requirements, while maintaining a high level of safety and reliability at a relatively low cost. The Heatpipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive space fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed-which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Heatpipe-Operated Mars Exploration Reactor (HOMER) is a derivative of the HPS designed especially for producing power on the surface of Mars. The HOMER-15 is a 15-kWt reactor that couples with a 3-kWe Stirling engine power system. The reactor contains stainless-steel (SS)-clad uranium nitride (UN) fuel pins that are structurally and thermally bonded to SS/sodium heatpipes. Fission energy is conducted from the fuel pins to the heatpipes, which then carry the heat to the Stirling engine. This paper describes the attributes, specifications, and performance of a 15-kWt HOMER reactor. .

Nuclear reactor control column

DOEpatents

Bachovchin, Dennis M.

1982-01-01

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

Methods of conducting simultaneous exothermic and endothermic reactions

DOEpatents

Tonkovich, Anna Lee [Marysville, OH; Roberts, Gary L [West Richland, WA; Perry, Steven T [Galloway, OH; Fitzgerald, Sean P [Columbus, OH

2005-11-29

Integrated Combustion Reactors (ICRs) and methods of making ICRs are described in which combustion chambers (or channels) are in direct thermal contact to reaction chambers for an endothermic reaction. Superior results were achieved for combustion chambers which contained a gap for free flow through the chamber. Particular reactor designs are also described. Processes of conducting reactions in integrated combustion reactors are described and results presented. Some of these processes are characterized by unexpected and superior results.

NEUTRONIC REACTOR CONTROL ELEMENT

DOEpatents

Newson, H.W.

1960-09-13

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

Automatic safety rod for reactors. [LMFBR

DOEpatents

Germer, J.H.

1982-03-23

An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

Automatic safety rod for reactors

DOEpatents

Germer, John H.

1988-01-01

An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

Process for treating effluent from a supercritical water oxidation reactor

DOEpatents

Barnes, C.M.; Shapiro, C.

1997-11-25

A method for treating a gaseous effluent from a supercritical water oxidation reactor containing entrained solids is provided comprising the steps of expanding the gas/solids effluent from a first to a second lower pressure at a temperature at which no liquid condenses; separating the solids from the gas effluent; neutralizing the effluent to remove any acid gases; condensing the effluent; and retaining the purified effluent to the supercritical water oxidation reactor. 6 figs.

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fanning, T. H.; Brunett, A. J.; Sumner, T.

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transientmore » thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.« less

Comparative testing of nondestructive examination techniques for concrete structures

NASA Astrophysics Data System (ADS)

Clayton, Dwight A.; Smith, Cyrus M.

2014-03-01

A multitude of concrete-based structures are typically part of a light water reactor (LWR) plant to provide foundation, support, shielding, and containment functions. Concrete has been used in the construction of nuclear power plants (NPPs) because of three primary properties, its inexpensiveness, its structural strength, and its ability to shield radiation. Examples of concrete structures important to the safety of LWR plants include containment building, spent fuel pool, and cooling towers. Comparative testing of the various NDE concrete measurement techniques requires concrete samples with known material properties, voids, internal microstructure flaws, and reinforcement locations. These samples can be artificially created under laboratory conditions where the various properties can be controlled. Other than NPPs, there are not many applications where critical concrete structures are as thick and reinforced. Therefore, there are not many industries other than the nuclear power plant or power plant industry that are interested in performing NDE on thick and reinforced concrete structures. This leads to the lack of readily available samples of thick and heavily reinforced concrete for performing NDE evaluations, research, and training. The industry that typically performs the most NDE on concrete structures is the bridge and roadway industry. While bridge and roadway structures are thinner and less reinforced, they have a good base of NDE research to support their field NDE programs to detect, identify, and repair concrete failures. This paper will summarize the initial comparative testing of two concrete samples with an emphasis on how these techniques could perform on NPP concrete structures.

10 CFR 100.2 - Scope.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Scope. 100.2 Section 100.2 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA § 100.2 Scope. The siting requirements contained in this part... and testing reactors pursuant to the provisions of part 50 or part 52 of this chapter. [61 FR 65175...

10 CFR 100.2 - Scope.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Scope. 100.2 Section 100.2 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA § 100.2 Scope. The siting requirements contained in this part... and testing reactors pursuant to the provisions of part 50 or part 52 of this chapter. [61 FR 65175...

CATALYTIC OXIDATION OF METHANE AT LOW SPACE VELOCITIES.

DTIC Science & Technology

methane in airstream through an inhouse designed and fabricated stainless steel reactor. The reactor contained either Hopcalite , 5% V2O5 - 5% MoO3 on...plotted for each catalyst flow rate combination and the effect of space velocity on conversion at constant temperature is shown for the Hopcalite and

Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H. Volume 2, Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Waganer, L.M.; Driemeyer, D.E.; Lee, V.D.

1992-03-01

This report contains a review of design studies for Inertial Confinement reactor. This second of three volumes discussions is some detail the following: Objectives, requirements, and assumptions; rationale for design option selection; key technical issues and R&D requirements; and conceptual design selection and description.

Decommissioning of eight surplus production reactors at the Hanford Site, Richland, Washington. Addendum (Final Environmental Impact Statement)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-12-01

The first section of this volume summarizes the content of the draft environmental impact statement (DEIS) and this Addendum, which together constitute the final environmental impact statement (FEIS) prepared on the decommissioning of eight surplus plutonium production reactors at Hanford. The FEIS consists of two volumes. The first volume is the DEIS as written. The second volume (this Addendum) consists of a summary; Chapter 9, which contains comments on the DEIS and provides DOE`s responses to the comments; Appendix F, which provides additional health effects information; Appendix K, which contains costs of decommissioning in 1990 dollars; Appendix L, which containsmore » additional graphite leaching data; Appendix M, which contains a discussion of accident scenarios; Appendix N, which contains errata; and Appendix 0, which contains reproductions of the letters, transcripts, and exhibits that constitute the record for the public comment period.« less

A Monte Carlo program to calculate the exposure rate from airborne radioactive gases inside a nuclear reactor containment building.

PubMed

Sherbini, S; Tamasanis, D; Sykes, J; Porter, S W

1986-12-01

A program was developed to calculate the exposure rate resulting from airborne gases inside a reactor containment building. The calculations were performed at the location of a wall-mounted area radiation monitor. The program uses Monte Carlo techniques and accounts for both the direct and scattered components of the radiation field at the detector. The scattered component was found to contribute about 30% of the total exposure rate at 50 keV and dropped to about 7% at 2000 keV. The results of the calculations were normalized to unit activity per unit volume of air in the containment. This allows the exposure rate readings of the area monitor to be used to estimate the airborne activity in containment in the early phases of an accident. Such estimates, coupled with containment leak rates, provide a method to obtain a release rate for use in offsite dose projection calculations.

FUEL ASSAY REACTOR

DOEpatents

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

1962-12-25

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

Fuel subassembly leak test chamber for a nuclear reactor

DOEpatents

Divona, Charles J.

1978-04-04

A container with a valve at one end is inserted into a nuclear reactor coolant pool. Once in the pool, the valve is opened by a mechanical linkage. An individual fuel subassembly is lifted into the container by a gripper; the valve is then closed providing an isolated chamber for the subassembly. A vacuum is drawn on the chamber to encourage gaseous fission product leakage through any defects in the cladding of the fuel rods comprising the subassembly; this leakage may be detected by instrumentation, and the need for replacement of the assembly ascertained.

Apparatus for controlling nuclear core debris

DOEpatents

Jones, Robert D.

1978-01-01

Nuclear reactor apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling.

METHOD AND APPARATUS FOR PRODUCING POWER

DOEpatents

Wollan, E.O.

1961-06-27

A neutronic reactor comprising two discrete zones; namely, an inner zone containing fissionable material and an outer zone containing fertile material is described. The inner zone is operated at a low temperature and is cooled by pressurized water. The outer zone is operated at a substantially higher temperature and is cooled by steam flashed from the inner zone. The reactor is particularly advantageous in that it produces high temperature steam; yet the materials of construction in the core (inner zone) are not restricted to materials capable of withstanding high temperature operation.

Results of using frequency banded SAFT for examining three types of defects

NASA Astrophysics Data System (ADS)

Clayton, Dwight; Barker, Alan; Santos-Villalobos, Hector

2017-02-01

A multitude of concrete-based structures are typically part of a light water reactor (LWR) plant to provide the foundation, support, shielding, and containment functions. Concrete has been used in the construction of nuclear power plants (NPPs) because of three primary properties; its low cost, structural strength, and ability to shield radiation. Examples of concrete structures important to the safety of LWR plants include the containment building, spent fuel pool, and cooling towers. This use has made concrete's long-term performance crucial for the safe operation of commercial NPPs. Extending reactor life to 60 years and beyond will likely increase susceptibility and severity of known forms of degradation. Additionally, new mechanisms of materials degradation are also possible. Specially designed and fabricated test specimens can provide realistic flaws that are similar to actual flaws in terms of how they interact with a particular Nondestructive Evaluation (NDE) technique. Artificial test blocks allow the isolation of certain testing problems as well as the variation of certain parameters. Because conditions in the laboratory are controlled, the number of unknown variables can be decreased, making it possible to focus on specific aspects, investigate them in detail, and gain further information on the capabilities and limitations of each method. To minimize artifacts caused by boundary effects, the dimensions of the specimens should not be too compact. In this paper, we apply the frequency banded Synthetic Aperture Focusing Technique (SAFT) technique to a 2.134 m × 2.134 m × 1.016 m concrete test specimen with twenty deliberately embedded defects. These twenty embedded defects simulate voids (honeycombs), delamination, and embedded organic construction debris. Using the time-frequency technique of wavelet packet decomposition and reconstruction, the spectral content of the signal can be divided into two resulting child nodes. The resulting two nodes can then also be divided into two child nodes with each child node containing half of the bandwidth (spectral content) of its parent node. This process can be repeated until the bandwidth of the child nodes is sufficiently small. Once the desired bandwidth has been obtained, the band limited signal can be analyzed using SAFT, enabling the visualization of reflectivity of a frequency band and that band's interaction with the contents of the concrete structure. This paper examines the benefits of using frequency banded SAFT.

Ab-initio study of C and O impurities in uranium nitride

NASA Astrophysics Data System (ADS)

Lopes, Denise Adorno; Claisse, Antoine; Olsson, Pär

2016-09-01

Uranium nitride (UN) has been considered a potential fuel for Generation IV (GEN-IV) nuclear reactors as well as a possible new fuel for Light Water Reactors (LWR), which would permit an extension of the fuel residence time in the reactor. Carbon and oxygen impurities play a key role in the UN microstructure, influencing important parameters such as creep, swelling, gas release under irradiation, compatibility with structural steel and coolants, and thermal stability. In this work, a systematic study of the electronic structure of UN containing C and O impurities using first-principles calculations by the Density Functional Theory (DFT) method is presented. In order to describe accurately the localized U 5f electrons, the DFT + U formalism was adopted. Moreover, to avoid convergence toward metastable states, the Occupation Matrix Control (OMC) methodology was applied. The incorporation of C and O in the N-vacancy is found to be energetically favorable. In addition, only for O, the incorporation in the interstitial position is energetically possible, showing some degree of solubility for this element in this site. The binding energies show that the pairs (Csbnd Nvac) and (Osbnd Nvac) interact much further than the other defects, which indicate the possible occurrence of vacancy drag phenomena and clustering of these impurities in grain boundaries, dislocations and free surfaces. The migration energy of an impurity by single N-vacancy show that C and O employ different paths during diffusion. Oxygen migration requires significantly lower energy than carbon. This fact is due to flexibility in the Usbnd O chemical bonds, which bend during the diffusion forming a pseudo UO2 coordination. On the other hand, C and N have a directional and inflexible chemical bond with uranium; always requiring the octahedral coordination. These findings provide detailed insight into how these impurities behave in the UN matrix, and can be of great interest for assisting the development of this new nuclear fuel for next-generation reactors.

Pressurized water reactor flow skirt apparatus

DOEpatents

Kielb, John F.; Schwirian, Richard E.; Lee, Naugab E.; Forsyth, David R.

2016-04-05

A pressurized water reactor vessel having a flow skirt formed from a perforated cylinder structure supported in the lower reactor vessel head at the outlet of the downcomer annulus, that channels the coolant flow through flow holes in the wall of the cylinder structure. The flow skirt is supported at a plurality of circumferentially spaced locations on the lower reactor vessel head that are not equally spaced or vertically aligned with the core barrel attachment points, and the flow skirt employs a unique arrangement of hole patterns that assure a substantially balanced pressure and flow of the coolant over the entire underside of the lower core support plate.

COOLED NEUTRONIC REACTOR

DOEpatents

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

1958-03-18

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

Additive Manufacturing of Catalyst Substrates for Steam-Methane Reforming

NASA Astrophysics Data System (ADS)

Kramer, Michelle; McKelvie, Millie; Watson, Matthew

2018-01-01

Steam-methane reforming is a highly endothermic reaction, which is carried out at temperatures up to 1100 °C and pressures up to 3000 kPa, typically with a Ni-based catalyst distributed over a substrate of discrete alumina pellets or beads. Standard pellet geometries (spheres, hollow cylinders) limit the degree of mass transfer between gaseous reactants and catalyst. Further, heat is supplied to the exterior of the reactor wall, and heat transfer is limited due to the nature of point contacts between the reactor wall and the substrate pellets. This limits the degree to which the process can be intensified, as well as limiting the diameter of the reactor wall. Additive manufacturing now gives us the capability to design structures with tailored heat and mass transfer properties, not only within the packed bed of the reactor, but also at the interface between the reactor wall and the packed bed. In this work, the use of additive manufacturing to produce monolithic-structured catalyst substrate models, made from acrylonitrile-butadiene-styrene, with enhanced conductive heat transfer is described. By integrating the reactor wall into the catalyst substrate structure, the effective thermal conductivity increased by 34% from 0.122 to 0.164 W/(m K).

Magnetless magnetic fusion

NASA Astrophysics Data System (ADS)

Beklemishev, A. D.; Tajima, T.

1994-02-01

The authors propose a concept of thermonuclear fusion reactor in which the plasma pressure is balanced by direct gas-wall interaction in a high-pressure vessel. The energy confinement is achieved by means of the self-contained toroidal magnetic configuration sustained by an external current drive or charged fusion products. This field structure causes the plasma pressure to decrease toward the inside of the discharge and thus it should be magnetohydrodynamically stable. The maximum size, temperature and density profiles of the reactor are estimated. An important feature of confinement physics is the thin layer of cold gas at the wall and the adjacent transitional region of dense arc-like plasma. The burning condition is determined by the balance between these nonmagnetized layers and the current-carrying plasma. They suggest several questions for future investigation, such as the thermal stability of the transition layer and the possibility of an effective heating and current drive behind the dense edge plasma. The main advantage of this scheme is the absence of strong external magnets and, consequently, potentially cheaper design and lower energy consumption.

Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.

2008-06-23

This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been mademore » at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics computer codes and methodologies applied to SSTAR. Another section of the report details the SSTAR safety design approach which is based upon defense-in-depth providing multiple levels of protection against the release of radioactive materials and how the inherent safety features of the lead coolant, nitride fuel, fast neutron spectrum core, pool vessel configuration, natural circulation, and containment meet or exceed the requirements for each level of protection. The report also includes recent results of a systematic analysis by LANL of data on corrosion of candidate cladding and structural material alloys of interest to SSTAR by LBE and Pb coolants; the data were taken from a new database on corrosion by liquid metal coolants created at LANL. The analysis methodology that considers penetration of an oxidation front into the alloy and dissolution of the trailing edge of the oxide into the coolant enables the long-term corrosion rate to be extracted from shorter-term corrosion data thereby enabling an evaluation of alloy performance over long core lifetimes (e.g., 30 years) that has heretofore not been possible. A number of candidate alloy specimens with special treatments or coatings which might enhance corrosion resistance at the temperatures at which SSTAR would operate were analyzed following testing in the DELTA loop at LANL including steels that were treated by laser peening at LLNL; laser peening is an approach that alters the oxide-metal bonds which could potentially improve corrosion resistance. LLNL is also carrying out Multi-Scale Modeling of the Fe-Cr system with the goal of assisting in the development of cladding and structural materials having greater resistance to irradiation.« less

Boiling water reactor radiation shielded Control Rod Drive Housing Supports

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baversten, B.; Linden, M.J.

1995-03-01

The Control Rod Drive (CRD) mechanisms are located in the area below the reactor vessel in a Boiling Water Reactor (BWR). Specifically, these CRDs are located between the bottom of the reactor vessel and above an interlocking structure of steel bars and rods, herein identified as CRD Housing Supports. The CRD Housing Supports are designed to limit the travel of a Control Rod and Control Rod Drive in the event that the CRD vessel attachement went to fail, allowing the CRD to be ejected from the vessel. By limiting the travel of the ejected CRD, the supports prevent a nuclearmore » overpower excursion that could occur as a result of the ejected CRD. The Housing Support structure must be disassembled in order to remove CRDs for replacement or maintenance. The disassembly task can require a significant amount of outage time and personnel radiation exposure dependent on the number and location of the CRDs to be changed out. This paper presents a way to minimize personal radiation exposure through the re-design of the Housing Support structure. The following paragraphs also delineate a method of avoiding the awkward, manual, handling of the structure under the reactor vessel during a CRD change out.« less

Concepts and Tests for the Remote-Controlled Dismantling of the Biological Shield and Form work of the KNK Reactor - 13425

DOE Office of Scientific and Technical Information (OSTI.GOV)

Neff, Sylvia; Graf, Anja; Petrick, Holger

The compact sodium-cooled nuclear reactor facility Karlsruhe (KNK), a prototype Fast Breeder, is currently in an advanced stage of dismantling. Complete dismantling is based on 10 partial licensing steps. In the frame of the 9. decommissioning permit, which is currently ongoing, the dismantling of the biological shield is foreseen. The biological shield consists of heavy reinforced concrete with built-in steel fitments, such as form-work of the reactor tank, pipe sleeves, ventilation channels, and measuring devices. Due to the activation of the inner part of the biological shield, dismantling has to be done remote-controlled. During a comprehensive basic design phase amore » practical dismantling strategy was developed. Necessary equipment and tools were defined. Preliminary tests revealed that hot wire plasma cutting is the most favorable cutting technology due to the geometrical boundary conditions, the varying distance between cutter and material, and the heavy concrete behind the steel form-work. The cutting devices will be operated remotely via a carrier system with an industrial manipulator. The carrier system has expandable claws to adjust to the varying diameter of the reactor shaft during dismantling progress. For design approval of this prototype development, interaction between manipulator and hot wire plasma cutting was tested in a real configuration. For the demolition of the concrete structure, an excavator with appropriate tools, such as a hydraulic hammer, was selected. Other mechanical cutting devices, such as a grinder or rope saw, were eliminated because of concrete containing steel spheres added to increase the shielding factor of the heavy concrete. Dismantling of the biological shield will be done in a ring-wise manner due to static reasons. During the demolition process, the excavator is positioned on its tripod in three concrete recesses made prior to the dismantling of the separate concrete rings. The excavator and the manipulator carrier system will be operated alternately. Main boundary condition for all the newly designed equipment is the decommissioning housing of limited space within the reactor building containment. To allow for a continuous removal of the concrete rubble, an additional opening on the lowest level of the reactor shaft will be made. All equipment and the interaction of the tools have to be tested before use in the controlled area. Therefore a full-scale model of the biological shield will be provided in a mock-up. The tests will be performed in early 2014. The dismantling of the biological shield is scheduled for 2015. (authors)« less

Illumina MiSeq sequencing reveals microbial community in HA process for dyeing wastewater treatment fed with different co-substrates.

PubMed

Xie, Xuehui; Liu, Na; Ping, Jing; Zhang, Qingyun; Zheng, Xiulin; Liu, Jianshe

2018-06-01

In present study, a hydrolysis acidification (HA) reactor was used for simulated dyeing wastewater treatment. Co-substrates included starch, glucose, sucrose, yeast extract (YE) and peptone were fed sequentially into the HA reactor to enhance the HA process effects. The performance of the HA reactor and the microbial community structure in HA process were investigated under different co-substrates conditions. Results showed that different co-substrates had different influences on the performance of HA reactor. The highest decolorization (50.64%) and COD removal rate (60.73%) of the HA reactor were obtained when sucrose was as the co-substrate. And it found that carbon co-substrates starch, glucose and sucrose exhibited better decolorization and higher COD removal efficiency of the HA reactor than the nitrogen co-substrates YE and peptone. Microbial community structure in the HA process was analyzed by Illumina MiSeq sequencing. Results revealed different co-substrates had different influences on the community structure and microbial diversity in HA process. It was considered that sucrose could enrich the species such as Raoultella, Desulfovibrio, Tolumonas, Clostridium, which might be capable of degrading the dyes. Sucrose was considered to be the best co-substrate of enhancing the HA reactor's performance in this study. This work would provide deep insight into the influence of many different co-substrates on HA reactor performance and microbial communities in HA process. Copyright © 2018 Elsevier Ltd. All rights reserved.

Synthesis of structured triacylglycerols containing caproic acid by lipase-catalyzed acidolysis: optimization by response surface methodology.

PubMed

Zhou, D; Xu, X; Mu, H; Høy, C E; Adler-Nissen, J

2001-12-01

Production in a batch reactor with a solvent-free system of structured triacylglycerols containing short-chain fatty acids by Lipozyme RM IM-catalyzed acidolysis between rapeseed oil and caproic acid was optimized using response surface methodology (RSM). Reaction time (t(r)), substrate ratio (S(r)), enzyme load (E(l), based on substrate), water content (W(c), based on enzyme), and reaction temperature (T(e)), the five most important parameters for the reaction, were chosen for the optimization. The range of each parameter was selected as follows: t(r) = 5-17 h; E(l) = 6-14 wt %; T(e) = 45-65 degrees C; S(r) = 2-6 mol/mol; and W(c) = 2-12 wt %. The biocatalyst was Lipozyme RM IM, in which Rhizomucor miehei lipase is immobilized on a resin. The incorporation of caproic acid into rapeseed oil was the main monitoring response. In addition, the contents of mono-incorporated structured triacylglycerols and di-incorporated structured triacylglycerols were also evaluated. The optimal reaction conditions for the incorporation of caproic acid and the content of di-incorporated structured triacylglycerols were as follows: t(r) = 17 h; S(r) = 5; E(l) = 14 wt %; W(c) = 10 wt %; T(e) = 65 degrees C. At these conditions, products with 55 mol % incorporation of caproic acid and 55 mol % di-incorporated structured triacylglycerols were obtained.

Liquid metal cooled nuclear reactor plant system

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1993-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

Combustion flame-plasma hybrid reactor systems, and chemical reactant sources

DOEpatents

Kong, Peter C

2013-11-26

Combustion flame-plasma hybrid reactor systems, chemical reactant sources, and related methods are disclosed. In one embodiment, a combustion flame-plasma hybrid reactor system comprising a reaction chamber, a combustion torch positioned to direct a flame into the reaction chamber, and one or more reactant feed assemblies configured to electrically energize at least one electrically conductive solid reactant structure to form a plasma and feed each electrically conductive solid reactant structure into the plasma to form at least one product is disclosed. In an additional embodiment, a chemical reactant source for a combustion flame-plasma hybrid reactor comprising an elongated electrically conductive reactant structure consisting essentially of at least one chemical reactant is disclosed. In further embodiments, methods of forming a chemical reactant source and methods of chemically converting at least one reactant into at least one product are disclosed.

Stimulation of the hydrolytic stage for biogas production from cattle manure in an electrochemical bioreactor.

PubMed

Samani, Saeed; Abdoli, Mohammad Ali; Karbassi, Abdolreza; Amin, Mohammad Mehdi

Electrical current in the hydrolytic phase of the biogas process might affect biogas yield. In this study, four 1,150 mL single membrane-less chamber electrochemical bioreactors, containing two parallel titanium plates were connected to the electrical source with voltages of 0, -0.5, -1 and -1.5 V, respectively. Reactor 1 with 0 V was considered as a control reactor. The trend of biogas production was precisely checked against pH, oxidation reduction potential and electrical power at a temperature of 37 ± 0.5°C amid cattle manure as substrate for 120 days. Biogas production increased by voltage applied to Reactors 2 and 3 when compared with the control reactor. In addition, the electricity in Reactors 2 and 3 caused more biogas production than Reactor 4. Acetogenic phase occurred more quickly in Reactor 3 than in the other reactors. The obtained results from Reactor 4 were indicative of acidogenic domination and its continuous behavior under electrical stimulation. The results of the present investigation clearly revealed that phasic electrical current could enhance the efficiency of biogas production.

Basic Nuclear Physics.

ERIC Educational Resources Information Center

Bureau of Naval Personnel, Washington, DC.

Basic concepts of nuclear structures, radiation, nuclear reactions, and health physics are presented in this text, prepared for naval officers. Applications to the area of nuclear power are described in connection with pressurized water reactors, experimental boiling water reactors, homogeneous reactor experiments, and experimental breeder…

Baseline Fracture Toughness and CGR testing of alloys X-750 and XM-19 (EPRI Phase I)

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. H. Jackson; S. P. Teysseyre

2012-10-01

The Advanced Test Reactor National Scientific User Facility (ATR NSUF) and Electric Power Research Institute (EPRI) formed an agreement to test representative alloys used as reactor structural materials as a pilot program toward establishing guidelines for future ATR NSUF research programs. This report contains results from the portion of this program established as Phase I (of three phases) that entails baseline fracture toughness, stress corrosion cracking (SCC), and tensile testing of selected materials for comparison to similar tests conducted at GE Global Research. The intent of this Phase I research program is to determine baseline properties for the materials ofmore » interest prior to irradiation, and to ensure comparability between laboratories using similar testing techniques, prior to applying these techniques to the same materials after having been irradiated at the Advanced Test Reactor (ATR). The materials chosen for this research are the nickel based super alloy X-750, and nitrogen strengthened austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co. and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research and provided to INL.« less

Baseline Fracture Toughness and CGR testing of alloys X-750 and XM-19 (EPRI Phase I)

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. H. Jackson; S. P. Teysseyre

2012-02-01

The Advanced Test Reactor National Scientific User Facility (ATR NSUF) and Electric Power Research Institute (EPRI) formed an agreement to test representative alloys used as reactor structural materials as a pilot program toward establishing guidelines for future ATR NSUF research programs. This report contains results from the portion of this program established as Phase I (of three phases) that entails baseline fracture toughness, stress corrosion cracking (SCC), and tensile testing of selected materials for comparison to similar tests conducted at GE Global Research. The intent of this Phase I research program is to determine baseline properties for the materials ofmore » interest prior to irradiation, and to ensure comparability between laboratories using similar testing techniques, prior to applying these techniques to the same materials after having been irradiated at the Advanced Test Reactor (ATR). The materials chosen for this research are the nickel based super alloy X-750, and nitrogen strengthened austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co. and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research and provided to INL.« less

Induction-heating MOCVD reactor with significantly improved heating efficiency and reduced harmful magnetic coupling

NASA Astrophysics Data System (ADS)

Li, Kuang-Hui; Alotaibi, Hamad S.; Sun, Haiding; Lin, Ronghui; Guo, Wenzhe; Torres-Castanedo, Carlos G.; Liu, Kaikai; Valdes-Galán, Sergio; Li, Xiaohang

2018-04-01

In a conventional induction-heating III-nitride metalorganic chemical vapor deposition (MOCVD) reactor, the induction coil is outside the chamber. Therefore, the magnetic field does not couple with the susceptor well, leading to compromised heating efficiency and harmful coupling with the gas inlet and thus possible overheating. Hence, the gas inlet has to be at a minimum distance away from the susceptor. Because of the elongated flow path, premature reactions can be more severe, particularly between Al- and B-containing precursors and NH3. Here, we propose a structure that can significantly improve the heating efficiency and allow the gas inlet to be closer to the susceptor. Specifically, the induction coil is designed to surround the vertical cylinder of a T-shaped susceptor comprising the cylinder and a top horizontal plate holding the wafer substrate within the reactor. Therefore, the cylinder coupled most magnetic field to serve as the thermal source for the plate. Furthermore, the plate can block and thus significantly reduce the uncoupled magnetic field above the susceptor, thereby allowing the gas inlet to be closer. The results show approximately 140% and 2.6 times increase in the heating and susceptor coupling efficiencies, respectively, as well as a 90% reduction in the harmful magnetic flux on the gas inlet.

Long-term effects of operating temperature and sulphate addition on the methanogenic community structure of anaerobic hybrid reactors.

PubMed

Pender, Seán; Toomey, Margaret; Carton, Micheál; Eardly, Dónal; Patching, John W; Colleran, Emer; O'Flaherty, Vincent

2004-02-01

The diversity, population dynamics, and activity profiles of methanogens in anaerobic granular sludges from two anaerobic hybrid reactors treating a molasses wastewater both mesophilically (37 degrees C) and thermophilically (55 degrees C) during a 1081 day trial were determined. The influent to one of the reactors was supplemented with sulphate, after an acclimation period of 112 days, to determine the effect of competition with sulphate-reducing bacteria on the methanogenic community structure. Sludge samples were removed from the reactors at intervals throughout the operational period and examined by amplified ribosomal DNA (rDNA) restriction analysis (ARDRA) and partial sequencing of 16S rRNA genes. In total, 18 operational taxonomic units (OTUs) were identified, 12 of which were sequenced. The methanogenic communities in both reactors changed during the operational period. The seed sludge and the reactor biomass sampled during mesophilic operation, both in the presence and absence of sulphate, was characterised by a predominance of Methanosaeta spp. Following temperature elevation, the dominant methanogenic sequences detected in the non-sulphate supplemented reactor were closely related to Methanocorpusculum parvum. By contrast, the dominant OTUs detected in the sulphate-supplemented reactor upon temperature increase were related to the hydrogen-utilising methanogen, Methanobacterium thermoautotrophicum. The observed methanogenic community structure in the reactors correlated with the operational performance of the reactors during the trial and with physiological measurements of the reactor biomass. Both reactors achieved chemical oxygen demand (COD) removal efficiencies of over 90% during mesophilic operation, with or without sulphate supplementation. During thermophilic operation, the presence of sulphate resulted in decreased reactor performance (effluent acetate concentrations of >3000 mg/l and biogas methane content of <25%). It was demonstrated that methanogenic conversion of acetate at 55 degrees C was extremely sensitive to inhibition by sulphide (50% inhibition at 8-17 mg/l unionised sulphide at pH 7.6-8.0), while the conversion of H(2)/CO(2) methanogenically was favoured. The combination of experiments carried out demonstrated the presence of specific methanogenic populations during periods of successful operational performance.

Application of two component biodegradable carriers in a particle-fixed biofilm airlift suspension reactor: development and structure of biofilms.

PubMed

Hille, Andrea; He, Mei; Ochmann, Clemens; Neu, Thomas R; Horn, Harald

2009-01-01

Two component biodegradable carriers for biofilm airlift suspension (BAS) reactors were investigated with respect to development of biofilm structure and oxygen transport inside the biofilm. The carriers were composed of PHB (polyhydroxybutyrate), which is easily degradable and PCL (caprolactone), which is less easily degradable by heterotrophic microorganisms. Cryosectioning combined with classical light microscopy and CLSM was used to identify the surface structure of the carrier material over a period of 250 days of biofilm cultivation in an airlift reactor. Pores of 50 to several hundred micrometers depth are formed due to the preferred degradation of PHB. Furthermore, microelectrode studies show the transport mechanism for different types of biofilm structures, which were generated under different substrate conditions. At high loading rates, the growth of a rather loosely structured biofilm with high penetration depths of oxygen was found. Strong changes of substrate concentration during fed-batch mode operation of the reactor enhance the growth of filamentous biofilms on the carriers. Mass transport in the outer regions of such biofilms was mainly driven by advection.

10 CFR 50.73 - Licensee event report system.

Code of Federal Regulations, 2013 CFR

2013-01-01

... scram or reactor trip. (2) General containment isolation signals affecting containment isolation valves... plant design; or (2) Normal and expected wear or degradation. (x) Any event that posed an actual threat...

10 CFR 50.73 - Licensee event report system.

Code of Federal Regulations, 2014 CFR

2014-01-01

... scram or reactor trip. (2) General containment isolation signals affecting containment isolation valves... plant design; or (2) Normal and expected wear or degradation. (x) Any event that posed an actual threat...

10 CFR 50.73 - Licensee event report system.

Code of Federal Regulations, 2012 CFR

2012-01-01

... scram or reactor trip. (2) General containment isolation signals affecting containment isolation valves... plant design; or (2) Normal and expected wear or degradation. (x) Any event that posed an actual threat...

Fuel transfer system

DOEpatents

Townsend, Harold E.; Barbanti, Giancarlo

1994-01-01

A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool.

Fuel transfer system

DOEpatents

Townsend, H.E.; Barbanti, G.

1994-03-01

A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool. 6 figures.

DISCHARGE DEVICE FOR RADIOACTIVE MATERIAL

DOEpatents

Ohlinger, L.A.

1958-09-23

A device is described fur unloading bodies of fissionable material from a neutronic reactor. It is comprised essentially of a wheeled flat car having a receptacle therein containing a liquid coolant fur receiving and cooling the fuel elements as they are discharged from the reactor, and a reciprocating plunger fur supporting the fuel element during discharge thereof prior to its being dropped into the coolant. The flat car is adapted to travel along the face of the reactor adjacent the discharge ends of the coolant tubes.

FUEL ASSEMBLY FOR A NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.

1958-04-29

A fuel assembly for a nuclear reactor of the type wherein liquid coolant is circulated through the core of the reactor in contact with the external surface of the fuel elements is described. In this design a plurality of parallel plates containing fissionable material are spaced about one-tenth of an inch apart and are supported between a pair of spaced parallel side members generally perpendicular to the plates. The plates all have a small continuous and equal curvature in the same direction between the side members.

Process for operating equilibrium controlled reactions

DOEpatents

Nataraj, Shankar; Carvill, Brian Thomas; Hufton, Jeffrey Raymond; Mayorga, Steven Gerard; Gaffney, Thomas Richard; Brzozowski, Jeffrey Richard

2001-01-01

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.

Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

NASA Astrophysics Data System (ADS)

Tsibulskiy, V. F.; Andrianova, E. A.; Davidenko, V. D.; Rodionova, E. V.; Tsibulskiy, S. V.

2017-12-01

A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium-plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

NEUTRONIC REACTOR DESIGN TO REDUCE NEUTRON LOSS

DOEpatents

Mills, F.T.

1961-05-01

A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall which is surrounded by successive layers of pure fertile material and fertile material having moderator. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. As the steel has a smaller capture cross-section for the fast neutrons, then greater numbers of the neutrons will pass into the blanket thereby increasing the over-all efficiency of the reactor.

Bioconversion reactor

DOEpatents

McCarty, Perry L.; Bachmann, Andre

1992-01-01

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

Apparatus for isotopic alteration of mercury vapor

DOEpatents

Grossman, Mark W.; George, William A.; Marcucci, Rudolph V.

1988-01-01

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

Thermomechanical analysis of fast-burst reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, J.D.

1994-08-01

Fast-burst reactors are designed to provide intense, short-duration pulses of neutrons. The fission reaction also produces extreme time-dependent heating of the nuclear fuel. An existing transient-dynamic finite element code was modified specifically to compute the time-dependent stresses and displacements due to thermal shock loads of reactors. Thermomechanical analysis was then applied to determine structural feasibility of various concepts for an EDNA-type reactor and to optimize the mechanical design of the new SPR III-M reactor.

Pressurized fluidized bed reactor and a method of operating the same

DOEpatents

Isaksson, J.

1996-02-20

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.

Pressurized fluidized bed reactor and a method of operating the same

DOEpatents

Isaksson, Juhani

1996-01-01

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

Kinetic modelling and microbial community assessment of anaerobic biphasic fixed film bioreactor treating distillery spent wash.

PubMed

Acharya, Bhavik K; Pathak, Hilor; Mohana, Sarayu; Shouche, Yogesh; Singh, Vasdev; Madamwar, Datta

2011-08-01

Anaerobic digestion, microbial community structure and kinetics were studied in a biphasic continuously fed, upflow anaerobic fixed film reactor treating high strength distillery wastewater. Treatment efficiency of the bioreactor was investigated at different hydraulic retention times (HRT) and organic loading rates (OLR 5-20 kg COD m⁻³ d⁻¹). Applying the modified Stover-Kincannon model to the reactor, the maximum removal rate constant (U(max)) and saturation value constant (K(B)) were found to be 2 kg m⁻³ d⁻¹ and 1.69 kg m⁻³ d⁻¹ respectively. Bacterial community structures of acidogenic and methanogenic reactors were assessed using culture-independent analyses. Sequencing of 16S rRNA genes exhibited a total of 123 distinct operational taxonomic units (OTUs) comprising 49 from acidogenic reactor and 74 (28 of eubacteria and 46 of archaea) from methanogenic reactor. The findings reveal the role of Lactobacillus sp. (Firmicutes) as dominant acid producing organisms in acidogenic reactor and Methanoculleus sp. (Euryarchaeotes) as foremost methanogens in methanogenic reactor. Copyright © 2011 Elsevier Ltd. All rights reserved.

Code qualification of structural materials for AFCI advanced recycling reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Natesan, K.; Li, M.; Majumdar, S.

2012-05-31

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Codemore » Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors. Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded in the past. The availability and additional needs for the key experimental facilities are summarized at the end of the report. Detailed information covered in each Chapter is given.« less

A Review on the Potential Use of Austenitic Stainless Steels in Nuclear Fusion Reactors

NASA Astrophysics Data System (ADS)

Şahin, Sümer; Übeyli, Mustafa

2008-12-01

Various engineering materials; austenitic stainless steels, ferritic/martensitic steels, vanadium alloys, refractory metals and composites have been suggested as candidate structural materials for nuclear fusion reactors. Among these structural materials, austenitic steels have an advantage of extensive technological database and lower cost compared to other non-ferrous candidates. Furthermore, they have also advantages of very good mechanical properties and fission operation experience. Moreover, modified austenitic stainless (Ni and Mo free) have relatively low residual radioactivity. Nevertheless, they can't withstand high neutron wall load which is required to get high power density in fusion reactors. On the other hand, a protective flowing liquid wall between plasma and solid first wall in these reactors can eliminate this restriction. This study presents an overview of austenitic stainless steels considered to be used in fusion reactors.

Simulator platform for fast reactor operation and safety technology demonstration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vilim, R. B.; Park, Y. S.; Grandy, C.

2012-07-30

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 responsemore » 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.« less

Effect of pentachlorophenol and chemical oxygen demand mass concentrations in influent on operational behaviors of upflow anaerobic sludge blanket (UASB) reactor.

PubMed

Shen, Dong-Sheng; He, Ruo; Liu, Xin-Wen; Long, Yan

2006-08-25

Upflow anaerobic sludge blanket (UASB) reactor that was seeded with anaerobic sludge acclimated to chlorophenols was used to investigate the feasibility of anaerobic biotreatment of synthetic wastewater containing pentachlorophenol (PCP) with additional sucrose as carbon source. Two sets of UASB reactors were operated at one time. But the seeded sludge for the two reactors was different and Reactor I was seeded with the sludge that was acclimated to PCP completely for half a year, and Reactor II was seeded with the mixed sludge that was acclimated for half a year to PCP, 4-CP, 3-CP or 2-CP, respectively. The degradation of PCP and the operation fee treating the wastewater are affected by the concentration of MEDS (microorganism easily degradable substrate). So the confirmation of the suitable ratio of [COD] and [PCP] was the key factor of treating the wastewater containing PCP economically and efficiently. During the experiment, the synthetic wastewater with 180.0 mg L(-1) PCP and 1250-10000 mg L(-1) COD could be treated steadily in the experimental Reactor I. The removal efficiency of PCP was more than 99.5% and the removal efficiency of COD was up to 90%. [PCP] (concentration of PCP) in effluent was less than 0.5 mg L(-1). [PCP] in influent could affect proper [COD] (concentration of COD) range in influent that was required for maintenance of steady running of the experimental reactor with a hydraulic retention time (HRT) from 20 to 22 h. [PCP] in influent would directly affect the necessary [COD] in influent when the UASB reactor ran normally and treated the wastewater containing PCP. When [PCP] was 100.4, 151.6 and 180.8 mg L(-1) in influent, respectively, [COD] in influent had to be controlled about 1250-7500, 2500-5000 and 5000 mg L(-1) to maintain the UASB reactor steady running normally and contemporarily ensure that [COD] and [PCP] in effluent were less than 300 and 0.5 mg L(-1), respectively. With the increase of [PCP] in influent, the range of variation of [COD] in influent endured by the UASB reactor was decreasing. The ratios of [COD] and [PCP] in influent could affect removal efficiency of PCP and COD, the concentration of total volatile fatty acids (VFA) in effluent, biogas quantity and methane content in biogas. [PCP] in influent was linearly or semi-logarithmically correlated to [COD] in effluent when [COD] in influent was 5750+/-250 mg L(-1), and so was the relationship between [COD] in influent and [PCP] in effluent when [PCP] in influent was 100.4 or 151.6 mg L(-1), less than the maximum permissible [PCP]. The sources of seeded sludge, the way of sludge acclimation and the characteristics of anaerobic sludge could all affect the UASB reactor capacity treating PCP. When [PCP] were less than 180.8 mg L(-1) for Reactor I and 151.6 mg L(-1) for Reactor II, the variation of [PCP] in influent had little effect on the UASB reactor volume gas production rate and substrate gas production rate. And [VFA] and pH value in effluent were affected a little. Volume biogas production rate and substrate biogas production rate of the UASB reactor were only affected by [COD] and loading rate in influent. But when [PCP] was more than 151.6 mg L(-1) for Reactor II, the biogas production fell quickly and was over 3 days later. [VFA] in effluent from Reactor II increased up to 2198.1 mg L(-1) quickly and the pH value fell to less than 7. Reactor II could not run normally. The component of VFA accumulated quickly was mainly acetate (above 50%). With [PCP] increased from 7.9 to 180.8 mg L(-1) gradually in influent, the methane content in biogas from Reactor II decreased from 70% to 60%, but the reactor could still run normally. Then as for Reactor II, the content of methane have fallen from 75% to 45% or so quickly. And Reactor II could not run steadily. So the conclusion could be drown that too high [PCP] in influent for UASB reactor mainly inhibited the activity of methane-producing bacteria cultures utilizing the acetate.

Neutronic reactor

DOEpatents

Wende, Charles W. J.

1976-08-17

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

ON-SITE ENGINEERING REPORT OF THE SLURRY-PHASE BIOLOGICAL REACTOR FOR PILOT-SCALE TESTING ON CONTAMINATED SOIL

EPA Science Inventory

The performance of pilot-scale bioslurry treatment on creosote-contaminated soil was evaluated. Five reactors containing 66 L of slurry (30% soil by weight), were operated in parallel. The soil was a sandy soil with minor gravel content. The pilot-scale phase utilized an inoculum...

75 FR 16517 - Dominion Nuclear Connecticut, Inc.; Millstone Power Station, Unit Nos 1, 2, and 3; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-01

.... Borchardt, NRC, to M. S. Fertel, Nuclear Energy Institute, ADAMS Accession No. ML091410309). The licensee's... effect. The facility consists of one boiling water reactor and two pressurized water reactors located in... public. The supplemental January 12, 2010, letter contains, as an attachment, an environmental assessment...

Performance of semi-continuous membrane bioreactor in biogas production from toxic feedstock containing D-Limonene.

PubMed

Wikandari, Rachma; Youngsukkasem, Supansa; Millati, Ria; Taherzadeh, Mohammad J

2014-10-01

A novel membrane bioreactor configuration containing both free and encased cells in a single reactor was proposed in this work. The reactor consisted of 120g/L of free cells and 120g/L of encased cells in a polyvinylidene fluoride membrane. Microcrystalline cellulose (Avicel) and d-Limonene were used as the models of substrate and inhibitor for biogas production, respectively. Different concentrations of d-Limonene i.e., 1, 5, and 10g/L were tested, and an experiment without the addition of d-Limonene was prepared as control. The digestion was performed in a semi-continuous thermophilic reactor for 75 days. The result showed that daily methane production in the reactor with the addition of 1g/L d-Limonene was similar to that of control. A lag phase was observed in the presence of 5g/L d-Limonene; however, after 10 days, the methane production increased and reached a similar production to that of the control after 15 days. Copyright © 2014 Elsevier Ltd. All rights reserved.

Experimental study on the instability of Pressure Balance Injection System (PBIS)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Okamoto, Koji; Teshima, Hideyuki; Madarame, Haruki

1996-06-01

The Passive Safety Reactor has been developed to reduce the construction cost and to improve the safety. Japan Atomic Energy Research institute (JAERI) proposed the System-Integrated Pressurized Water Reactor (SPWR) as a Passive Safety Reactor. In the SPWR design, the Pressure Balanced Injection System (PBIS) was introduced for the passive safety concept. The water with boron in a containment vessel were passively injected into the core by the pressure difference between the containment vessel and reactor vessel at a severe accidental condition. However there are few studies on the thermo-hydraulic characteristics of the PBIS. In this study, the thermal hydraulicsmore » of the PBIS are experimentally investigated using the small scale model. The instability of the injected flow was observed in the adiabatic experiment. The instability was caused by the pressure balance between the two vessels. The mechanism of the instability are discussed, resulting in the good agreement with the experimental results. In the steam experiment, another instability was observed, which was caused by the heat balance in the main tank.« less

Energy efficient continuous flow ash lockhopper

NASA Technical Reports Server (NTRS)

Collins, Earl R., Jr. (Inventor); Suitor, Jerry W. (Inventor); Dubis, David (Inventor)

1989-01-01

The invention relates to an energy efficient continuous flow ash lockhopper, or other lockhopper for reactor product or byproduct. The invention includes an ash hopper at the outlet of a high temperature, high pressure reactor vessel containing heated high pressure gas, a fluidics control chamber having an input port connected to the ash hopper's output port and an output port connected to the input port of a pressure letdown means, and a control fluid supply for regulating the pressure in the control chamber to be equal to or greater than the internal gas pressure of the reactor vessel, whereby the reactor gas is contained while ash is permitted to continuously flow from the ash hopper's output port, impelled by gravity. The main novelty resides in the use of a control chamber to so control pressure under the lockhopper that gases will not exit from the reactor vessel, and to also regulate the ash flow rate. There is also novelty in the design of the ash lockhopper shown in two figures. The novelty there is the use of annular passages of progressively greater diameter, and rotating the center parts on a shaft, with the center part of each slightly offset from adjacent ones to better assure ash flow through the opening.

Recycling of hazardous solid waste material using high-temperature solar process heat. 2. Reactor design and experimentation.

PubMed

Schaffner, Beatrice; Meier, Anton; Wuillemin, Daniel; Hoffelner, Wolfgang; Steinfeld, Aldo

2003-01-01

A novel high-temperature solar chemical reactor is proposed for the thermal recycling of hazardous solid waste material using concentrated solar power. It features two cavities in series, with the inner one functioning as the solar absorber and the outer one functioning as the reaction chamber. The solar reactor can handle thermochemical processes at temperatures above 1,300 K involving multiphases and controlled atmospheres. It further allows for batch or continuous mode of operation and for easy adjustment of the residence time of the reactants to match the kinetics of the reaction. A 10-kW solar reactor prototype was designed and tested for the carbothermic reduction of electric arc furnace dusts (EAFD). The reactor was subjected to mean solar flux intensities of 2,000 kW m(-2) and operated in both batch and continuous mode within the temperature range of 1,120-1,400 K. Extraction of over 90% of the toxic compounds originally contained in the EAFD was achieved while the condensable products of the off-gas contained mainly Zn, Pb, and Cl. The use of concentrated solar energy as the source of process heat offers the possibility of converting hazardous solid waste material into valuable commodities for processes in closed and sustainable material cycles.

Biogas production from Jatropha curcas press-cake.

PubMed

Staubmann, R; Foidl, G; Foidl, N; Gübitz, G M; Lafferty, R M; Arbizu, V M; Steiner, W

1997-01-01

Seeds of the tropical plant Jatropha curcas (purge nut, physic nut) are used for the production of oil. Several methods for oil extraction have been developed. In all processes, about 50% of the weight of the seeds remain as a press cake containing mainly protein and carbohydrates. Investigations have shown that this residue contains toxic compounds and cannot be used as animal feed without further processing. Preliminary experiments have shown that the residue is a good substrate for biogas production. Biogas formation was studied using a semicontinous upflow anaerobic sludge blanket (UASB) reactor; a contact-process and an anaerobic filter each reactor having a total volume of 110 L. A maximum production rate of 3.5 m3 m"3 d"1 was obtained in the anaerobic filter with a loading rate of 13 kg COD m~3 d"1. However, the UASB reactor and the contact-process were not suitable for using this substrate. When using an anaerobic filter with Jatropha curcas seed cake as a substrate, 76% of the COD was degraded and 1 kg degraded COD yielded 355 L of biogas containing 70% methane.

Chapter 5, "License Renewal and Aging Management for Continued Service

DOE Office of Scientific and Technical Information (OSTI.GOV)

Naus, Dan J

As of August 2011, there were 104 commercial nuclear power reactors licensed to operate in 31 states in the United States. Initial operating licenses in the United States are granted for a period of 40 years. In order to help assure an adequate energy supply, the USNRC has established a timely license renewal process and clear requirements that are needed to ensure safe plant operation for an extended plant life. The principals of license renewal and the basic requirements that address license renewal are identified as well as additional sources of guidance that can be utilized as part of themore » license renewal process. Aging management program inspections and operating experience related to the concrete and steel containment structures are provided. Finally, several lessons learned are provided based on containment operating experience.« less

FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Davidson, J.K.

1963-11-19

A fuel element structure particularly useful in high temperature nuclear reactors is presented. Basically, the structure comprises two coaxial graphite sleeves integrally joined together by radial fins. Due to the high structural strength of graphite at high temperatures and the rigidity of this structure, nuclear fuel encased within the inner sleeve in contiguous relation therewith is supported and prevented from expanding radially at high temperatures. Thus, the necessity of relying on the usual cladding materials with relatively low temperature limitations for structural strength is removed. (AEC)

Hanging core support system for a nuclear reactor

DOEpatents

Burelbach, James P.; Kann, William J.; Pan, Yen-Cheng; Saiveau, James G.; Seidensticker, Ralph W.

1987-01-01

For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform. Motion or radiation sensing detectors can be provide at the lower ends of the tension rods for obtaining pertinent readings proximate the core.

Effects of intermittent aeration periods on a structured-bed reactor continuously fed on the post-treatment of sewage anaerobic effluent.

PubMed

Silva, Bruno Garcia; Damianovic, Márcia Helena Rissato Zamariolli; Foresti, Eugenio

2018-04-20

This study assessed the simultaneous nitrification and denitrification processes and remaining organic matter removal from anaerobic reactor effluent treating wastewater in a single reactor. A structured-bed reactor, with polyurethane foam as support media, was subjected to intermittent aeration and effluent recirculation. Aerated/non-aerated periods varied in the range of 2/1-1/3 h. The chemical oxygen demand (COD) in the effluent remained between 26 and 42 mg L -1 throughout all the aeration conditions. Aeration periods of 1/2 h removed 80 and 26% of Total Kjeldahl Nitrogen and Total Nitrogen, respectively. A low solid production was observed during the 300 days of operation, resulting in a solid retention time of 139 days. The results indicate that the non-aerated periods generated alkalinity that favored nitrification, maintaining low COD concentrations in the effluent. The structured bed reactor presented a low solid production and effluent loss below 20 mgSSV L -1 , similar to concentrations obtained in secondary decanters.

Optimization of hydraulic shear parameters and reactor configuration in the aerobic granular sludge process.

PubMed

Zhu, Liang; Zhou, Jiaheng; Yu, Haitian; Xu, Xiangyang

2015-01-01

The hydraulic shear acts as an important selection pressure in aerobic sludge granulation. The effects of the hydraulic shear rate and reactor configuration on structural characteristics of aerobic granule in view of the hydromechanics. The hydraulic shear analysis was proposed to overcome the limitation of using superficial gas velocity (SGV) to express the hydraulic shear stress. Results showed that the stronger hydraulic shear stress with SGV above 2.4 cm s(-1) promoted the microbial aggregation, and favoured the structural stability of the granular sludge. According to the hydraulic shear analysis, the total shear rate reached (0.56-2.31)×10(5) s(-1) in the granular reactor with a larger ratio of height to diameter (H/D), and was higher than that in the reactor with smaller H/D, where the sequencing airlift bioreactor with smaller H/D had a high total shear rate under the same SGV. Results demonstrated that the granular reactor could provide a stronger hydraulic shear stress which promotes the formation and structural stability of aerobic granules.

Treatment of synthetic refinery wastewater in anoxic-aerobic sequential moving bed reactors and sulphur recovery.

PubMed

Mallick, Subrat Kumar; Chakraborty, Saswati

2017-11-10

Objective of the present study was to simultaneously biodegrade synthetic petroleum refinery wastewater containing phenol (750 mg/L), sulphide (750 mg/L), hydrocarbon (as emulsified diesel of 300 mg/L), ammonia-nitrogen (350 mg/L) at pH >9 in anoxic-aerobic sequential moving bed reactors. The optimum mixing speed of anoxic reactor was observed at 20 rpm and beyond that, removal rate remained constant. In anoxic reactor the minimum hydraulic retention time was observed to be 2 days for complete removal of sulphide, 40-50% removal of phenol and total hydrocarbons and 52% of sulphur recovery. The optimum HRT of aerobic moving bed reactor was observed as 16 h (total HRT of 64 h for anoxic and aerobic reactors) for complete removals of phenol, total hydrocarbons, COD (chemical oxygen demand) and ammonia-nitrogen with nitrification.

Nuclear reactors built, being built, or planned 1996

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1997-08-01

This publication contains unclassified information about facilities, built, being built, or planned in the United States for domestic use or export as of December 31, 1996. The Office of Scientific and Technical Information, U.S. Department of Energy, gathers this information annually from Washington headquarters, and field offices of DOE; from the U.S. Nuclear Regulatory Commission (NRC); from the U. S. reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from U.S. and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables ofmore » the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled.« less

Catalytic wet oxidation: mathematical modeling of multicompound destruction.

PubMed

Yang, J; Hand, D W; Hokanson, D R; Crittenden, J C; Oman, E J

2003-01-01

A mathematical model of a three-phase catalytic reactor, CatReac, was developed for analysis and optimization of a catalytic oxidation reactor that is used in the International Space Station potable water processor. The packed-bed catalytic reactor, known as the volatile reactor assembly (VRA), is operated as a three-phase reactor and contains a proprietary catalyst, a pure-oxygen gas phase, and the contaminated water. The contaminated water being fed to the VRA primarily consists of acetic acid, acetone, ethanol, 1-propanol, 2-propanol, and propionic acid ranging in concentration from 1 to 10 mg/L. The Langmuir-Hinshelwood Hougen-Watson (L-H) (Hougen, 1943) expression was used to describe the surface reaction rate for these compounds. Single and multicompound short-column experiments were used to determine the L-H rate parameters and calibrate the model. The model was able to predict steady-state multicomponent effluent profiles for short and full-scale reactor experiments.

Simultaneous saccharification and extractive fermentation of lignocellulosic materials into lactic acid in a two-zone fermentor-extractor system.

PubMed

Iyer, P V; Lee, Y Y

1999-01-01

Simultaneous saccharification and extractive fermentation of lignocellulosic materials into lactic acid was investigated using a two-zone bioreactor. The system is composed of an immobilized cell reactor, a separate column reactor containing the lignocellulosic substrate and a hollow-fiber membrane. It is operated by recirculating the cell free enzyme (cellulase) solution from the immobilized cell reactor to the column reactor through the membrane. The enzyme and microbial reactions thus occur at separate locations, yet simultaneously. This design provides flexibility in reactor operation as it allows easy separation of the solid substrate from the microorganism, in situ removal of the product and, if desired, different temperatures in the two reactor sections. This reactor system was tested using pretreated switchgrass as the substrate. It was operated under a fed-batch mode with continuous removal of lactic acid by solvent extraction. The overall lactic acid yield obtainable from this bioreactor system is 77% of the theoretical.

Modeling the Homogenization Kinetics of As-Cast U-10wt% Mo alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, Zhijie; Joshi, Vineet; Hu, Shenyang Y.

2016-01-15

Low-enriched U-22at% Mo (U-10Mo) alloy has been considered as an alternative material to replace the highly enriched fuels in research reactors. For the U-10Mo to work effectively and replace the existing fuel material, a thorough understanding of the microstructure development from as-cast to the final formed structure is required. The as-cast microstructure typically resembles an inhomogeneous microstructure with regions containing molybdenum-rich and -lean regions, which may affect the processing and possibly the in-reactor performance. This as-cast structure must be homogenized by thermal treatment to produce a uniform Mo distribution. The development of a modeling capability will improve the understanding ofmore » the effect of initial microstructures on the Mo homogenization kinetics. In the current work, we investigated the effect of as-cast microstructure on the homogenization kinetics. The kinetics of the homogenization was modeled based on a rigorous algorithm that relates the line scan data of Mo concentration to the gray scale in energy dispersive spectroscopy images, which was used to generate a reconstructed Mo concentration map. The map was then used as realistic microstructure input for physics-based homogenization models, where the entire homogenization kinetics can be simulated and validated against the available experiment data at different homogenization times and temperatures.« less

Chemical compatibility of some ceramic matrix composite structures with fusion reactor helium coolant at high temperatures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Perez, F.J.; Ghoniem, N.M.

The thermodynamic stability of SiC/SiC composite structures proposed for fusion applications is presented in this paper. Minimization of the free energy for reacting species in the temperature range 773-1273 K is achieved by utilizing the NASA-Lewis Chemical Equilibrium Thermodynamics Code (CET). The chemical stability of the matrix (SiC), as well as several potential fiber coatings are studied. Helium coolant is assumed to contain O{sub 2} and water moisture impurities in the range 100-1000 ppm. The work is applied to recent Magnetic and Inertial Confinement Conceptual designs. The present study indicated that the upper useful temperature limit for SiC/SiC composites, frommore » the standpoint of high-temperature corrosion, will be in the neighborhood of 1273 K. Up to this temperature, corrosion of SiC is shown to be negligible. The main mechanism of weight loss will be by evaporation to the plasma side. The presence of a protective SiO{sub 2} condensed phase is discussed, and is shown to result in further reduction of high-temperature corrosion. The thermodynamic stability of C and BN is shown to be very poor under typical fusion reactor conditions. Further development of chemically stable interface materials is required.« less

Nitrification at different salinities: Biofilm community composition and physiological plasticity.

PubMed

Gonzalez-Silva, Blanca M; Jonassen, Kjell Rune; Bakke, Ingrid; Østgaard, Kjetill; Vadstein, Olav

2016-05-15

This paper describes an experimental study of microbial communities of three moving bed biofilm reactors (MBBR) inoculated with nitrifying cultures originated from environments with different salinity; freshwater, brackish (20‰) and seawater. All reactors were run until they operated at a conversion efficiency of >96%. The microbial communities were profiled using 454-pyrosequencing of 16S rRNA gene amplicons. Statistical analysis was used to investigate the differences in microbial community structure and distribution of the nitrifying populations with different salinity environments. Nonmetric multidimensional scaling analysis (NMDS) and the PERMANOVA test based on Bray-Curtis similarities revealed significantly different community structure in the three reactors. The brackish reactor showed lower diversity index than fresh and seawater reactors. Venn diagram showed that 60 and 78% of the total operational taxonomic units (OTUs) in the ammonia-oxidizing bacteria (AOB) and nitrite-oxidizing bacteria (NOB) guild, respectively, were unique OTUs for a given reactor. Similarity Percentages (SIMPER) analysis showed that two-thirds of the total difference in community structure between the reactors was explained by 10 OTUs, indicating that only a small number of OTUs play a numerically dominant role in the nitrification process. Acute toxicity of salt stress on ammonium and nitrite oxidizing activities showed distinctly different patterns, reaching 97% inhibition of the freshwater reactor for ammonium oxidation rate. In the brackish culture, inhibition was only observed at maximal level of salinity, 32‰. In the fully adapted seawater culture, higher activities were observed at 32‰ than at any of the lower salinities. Copyright © 2016 Elsevier Ltd. All rights reserved.

Characteristics of Pyrolytic Topping in Fluidized Bed for Different Volatile Coals

NASA Astrophysics Data System (ADS)

Xiong, R.; Dong, L.; Xu, G. W.

Coal is generally combusted or gasified directly to destroy completely the chemical structures, such as aromatic rings containing in volatile coals including bituminite and lignite. Coal topping refers to a process that extracts chemicals with aromatic rings from such volatile coals in advance of combustion or gasification and thereby takes advantage of the value of coal as a kind of chemical structure resource. CFB boiler is the coal utilization facility that can be easily retrofitted to implement coal topping. A critical issue for performing coal topping is the choice of the pyrolytic reactor that can be different types. The present study concerns fluidized bed reactor that has rarely been tested for use in coal topping. Two different types of coals, one being Xiaolongtan (XLT) lignite and the other Shanxi (SX) bituminous, were tested to clarify the yield and composition of pyrolysis liquid and gas under conditions simulating actual operations. The results showed that XLT lignite coals had the maximum tar yield in 823-873K and SX bituminite realized its highest tar yield in 873-923K. Overall, lignite produced lower tar yield than bituminous coal. The pyrolysis gas from lignite coals contained more CO and CO2 and less CH4, H2 and C2+C3 (C2H4, C2H6, C3H6, C3H8) components comparing to that from bituminous coal. TG-FTIR analysis of tars demonstrated that for different coals there are different amounts of typical chemical species. Using coal ash of CFB boiler, instead of quartz sand, as the fluidized particles decreased the yields of both tar and gas for all the tested coals. Besides, pyrolysis in a reaction atmosphere simulating the pyrolysis gas (instead of N2) resulted also in higher production of pyrolysis liquid.