F Reactor Inspection

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

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-10-29

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

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

Nuclear reactor construction with bottom supported reactor vessel

DOEpatents

Sharbaugh, John E.

1987-01-01

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

Structural response of 1/20-scale models of the Clinch River Breeder Reactor to a simulated hypothetical core disruptive accident. Technical report 4

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

Romander, C. M.; Cagliostro, D. J.

Five experiments were performed to help evaluate the structural integrity of the reactor vessel and head design and to verify code predictions. In the first experiment (SM 1), a detailed model of the head was loaded statically to determine its stiffness. In the remaining four experiments (SM 2 to SM 5), models of the vessel and head were loaded dynamically under a simulated 661 MW-sec hypothetical core disruptive accident (HCDA). Models SM 2 to SM 4, each of increasing complexity, systematically showed the effects of upper internals structures, a thermal liner, core support platform, and torospherical bottom on vessel response.more » Model SM 5, identical to SM 4 but more heavily instrumented, demonstrated experimental reproducibility and provided more comprehensive data. The models consisted of a Ni 200 vessel and core barrel, a head with shielding and simulated component masses, an upper internals structure (UIS), and, in the more complex models SM 4 and SM 5, a Ni 200 thermal liner and core support structure. Water simulated the liquid sodium coolant and a low-density explosive simulated the HCDA loads.« less

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

Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems.

Characterization of Sodium Thermal Hydraulics with Optical Fiber Temperature Sensors

NASA Astrophysics Data System (ADS)

Weathered, Matthew Thomas

The thermal hydraulic properties of liquid sodium make it an attractive coolant for use in Generation IV reactors. The liquid metal's high thermal conductivity and low Prandtl number increases efficiency in heat transfer at fuel rods and heat exchangers, but can also cause features such as high magnitude temperature oscillations and gradients in the coolant. Currently, there exists a knowledge gap in the mechanisms which may create these features and their effect on mechanical structures in a sodium fast reactor. Two of these mechanisms include thermal striping and thermal stratification. Thermal striping is the oscillating temperature field created by the turbulent mixing of non-isothermal flows. Usually this occurs at the reactor core outlet or in piping junctions and can cause thermal fatigue in mechanical structures. Meanwhile, thermal stratification results from large volumes of non-isothermal sodium in a pool type reactor, usually caused by a loss of coolant flow accident. This stratification creates buoyancy driven flow transients and high temperature gradients which can also lead to thermal fatigue in reactor structures. In order to study these phenomena in sodium, a novel method for the deployment of optical fiber temperature sensors was developed. This method promotes rapid thermal response time and high spatial temperature resolution in the fluid. The thermal striping and stratification behavior in sodium may be experimentally analyzed with these sensors with greater fidelity than ever before. Thermal striping behavior at a junction of non-isothermal sodium was fully characterized with optical fibers. An experimental vessel was hydrodynamically scaled to model thermal stratification in a prototypical sodium reactor pool. Novel auxiliary applications of the optical fiber temperature sensors were developed throughout the course of this work. One such application includes local convection coefficient determination in a vessel with the corollary application of level sensing. Other applications were cross correlation velocimetry to determine bulk sodium flow rate and the characterization of coherent vortical structures in sodium with temperature frequency data. The data harvested, instrumentation developed and techniques refined in this work will help in the design of more robust reactors as well as validate computational models for licensing sodium fast reactors.

Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

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

Bardet, Philippe; Ricciardi, Guillaume

2016-01-31

Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWRmore » fuel bundle behavior during seismic transients.« less

Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

DOE PAGES

Christon, Mark A.; Lu, Roger; Bakosi, Jozsef; ...

2016-10-01

Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuelmore » rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid–structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Furthermore, robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.« less

Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

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

Christon, Mark A.; Lu, Roger; Bakosi, Jozsef

Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuelmore » rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid–structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Furthermore, robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.« less

Thermionic switched self-actuating reactor shutdown system

DOEpatents

Barrus, Donald M.; Shires, Charles D.; Brummond, William A.

1989-01-01

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

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.

Assessing thermal conductivity of composting reactor with attention on varying thermal resistance between compost and the inner surface.

PubMed

Wang, Yongjiang; Niu, Wenjuan; Ai, Ping

2016-12-01

Dynamic estimation of heat transfer through composting reactor wall was crucial for insulating design and maintaining a sanitary temperature. A model, incorporating conductive, convective and radiative heat transfer mechanisms, was developed in this paper to provide thermal resistance calculations for composting reactor wall. The mechanism of thermal transfer from compost to inner surface of structural layer, as a first step of heat loss, was important for improving insulation performance, which was divided into conduction and convection and discussed specifically in this study. It was found decreasing conductive resistance was responsible for the drop of insulation between compost and reactor wall. Increasing compost porosity or manufacturing a curved surface, decreasing the contact area of compost and the reactor wall, might improve the insulation performance. Upon modeling of heat transfers from compost to ambient environment, the study yielded a condensed and simplified model that could be used to conduct thermal resistance analysis for composting reactor. With theoretical derivations and a case application, the model was applicable for both dynamic estimation and typical composting scenario. Copyright © 2016 Elsevier Ltd. All rights reserved.

Response of a continuous anaerobic digester to temperature transitions: A critical range for restructuring the microbial community structure and function.

PubMed

Kim, Jaai; Lee, Changsoo

2016-02-01

Temperature is a crucial factor that significantly influences the microbial activity and so the methanation performance of an anaerobic digestion (AD) process. Therefore, how to control the operating temperature for optimal activity of the microbes involved is a key to stable AD. This study examined the response of a continuous anaerobic reactor to a series of temperature shifts over a wide range of 35-65 °C using a dairy-processing byproduct as model wastewater. During the long-term experiment for approximately 16 months, the reactor was subjected to stepwise temperature increases by 5 °C at a fixed HRT of 15 days. The reactor showed stable performance within the temperature range of 35-45 °C, with the methane production rate and yield being maximum at 45 °C (18% and 26% greater, respectively, than at 35 °C). However, the subsequent increase to 50 °C induced a sudden performance deterioration with a complete cessation of methane recovery, indicating that the temperature range between 45 °C and 50 °C had a critical impact on the transition of the reactor's methanogenic activity from mesophilic to thermophilic. This serious process perturbation was associated with a severe restructuring of the reactor microbial community structure, particularly of methanogens, quantitatively as well as qualitatively. Once restored by interrupted feeding for about two months, the reactor maintained fairly stable performance under thermophilic conditions until it was upset again at 65 °C. Interestingly, in contrast to most previous reports, hydrogenotrophs largely dominated the methanogen community at mesophilic temperatures while acetotrophs emerged as a major group at thermophilic temperature. This implies that the primary methanogenesis route of the reactor shifted from hydrogen- to acetate-utilizing pathways with the temperature shifts from mesophilic to thermophilic temperatures. Our observations suggest that a mesophilic digester may not need to be cooled at up to 45 °C in case of undesired temperature rise, for example, by excessive self-heating, which offers a possibility to reduce operating costs. Copyright © 2015 Elsevier Ltd. All rights reserved.

Nozzle seal

DOEpatents

Groff, Russell Dennis; Vatovec, Richard John

1978-06-11

In an illustrative embodiment of the invention, a nuclear reactor pressure vessel, having an internal hoop from which the heated coolant emerges from the reactor core and passes through to the reactor outlet nozzles, is provided with annular sealing members operatively disposed between the outlet nozzle and the hoop and partly within a retaining annulus formed in the hoop. The sealing members are biased against the pressure vessel and the hoop and one of the sealing members is provided with a piston type pressure ring sealing member which effectively closes the path between the inlet and outlet coolants in the region about the outlet nozzle establishing a leak-proof condition. Furthermore, the flexible responsiveness of the seal assures that the seal will not structurally couple the hoop to the pressure vessel.

Mechanical property changes induced in structural alloys by neutron irradiations with different helium to displacement ratios*1

NASA Astrophysics Data System (ADS)

Mansur, L. K.; Grossbeck, M. L.

1988-07-01

Effects of helium on mechanical properties of irradiated structural materials are reviewed. In particular, variations in response to the ratio of helium to displacement damage serve as the focus. Ductility in creep and tensile tests is emphasized. A variety of early work has led to the current concentration on helium effects for fusion reactor materials applications. A battery of techniques has been developed by which the helium to displacement ratio can be varied. Our main discussion is devoted to the techniques of spectral tailoring and isotopic alloying currently of interest for mixed-spectrum reactors. Theoretical models of physical mechanisms by which helium interacts with displacement damage have been developed in terms of hardening to dislocation motion and grain boundary cavitation. Austenitic stainless steels, ferritic/martensitic steels and vanadium alloys are considered. In each case, work at low strain rates, where the main problems may lie, at the helium to displacement ratios appropriate to fusion reactor materials is lacking. Recent experimental evidence suggests that both in-reactor and high helium results may differ substantially from post-irradiation or low helium results. It is suggested that work in these areas is especially needed.

Data report of a pretest analysis of soil-structure interaction and structural response in low-amplitude explosive testing (50 KG) of the heissdampfreaktor (HDR)

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

Vaughan, D.K.; Sandler, I.; Rubin, D.

This report describes a three-dimensional nonlinear TRANAL finite element analysis of a nuclear reactor subjected to ground shaking from a buried 50 kg explosive source. The analysis is a pretest simulation of a test event which was scheduled to be conducted in West Germany on 3 November 1979.

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.

Application of USNRC NUREG/CR-6661 and draft DG-1108 to evolutionary and advanced reactor designs

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

Chang 'Apollo', Chen

2006-07-01

For the seismic design of evolutionary and advanced nuclear reactor power plants, there are definite financial advantages in the application of USNRC NUREG/CR-6661 and draft Regulatory Guide DG-1108. NUREG/CR-6661, 'Benchmark Program for the Evaluation of Methods to Analyze Non-Classically Damped Coupled Systems', was by Brookhaven National Laboratory (BNL) for the USNRC, and Draft Regulatory Guide DG-1108 is the proposed revision to the current Regulatory Guide (RG) 1.92, Revision 1, 'Combining Modal Responses and Spatial Components in Seismic Response Analysis'. The draft Regulatory Guide DG-1108 is available at http://members.cox.net/apolloconsulting, which also provides a link to the USNRC ADAMS site to searchmore » for NUREG/CR-6661 in text file or image file. The draft Regulatory Guide DG-1108 removes unnecessary conservatism in the modal combinations for closely spaced modes in seismic response spectrum analysis. Its application will be very helpful in coupled seismic analysis for structures and heavy equipment to reduce seismic responses and in piping system seismic design. In the NUREG/CR-6661 benchmark program, which investigated coupled seismic analysis of structures and equipment or piping systems with different damping values, three of the four participants applied the complex mode solution method to handle different damping values for structures, equipment, and piping systems. The fourth participant applied the classical normal mode method with equivalent weighted damping values to handle differences in structural, equipment, and piping system damping values. Coupled analysis will reduce the equipment responses when equipment, or piping system and structure are in or close to resonance. However, this reduction in responses occurs only if the realistic DG-1108 modal response combination method is applied, because closely spaced modes will be produced when structure and equipment or piping systems are in or close to resonance. Otherwise, the conservatism in the current Regulatory Guide 1.92, Revision 1, will overshadow the advantage of coupled analysis. All four participants applied the realistic modal combination method of DG-1108. Consequently, more realistic and reduced responses were obtained. (authors)« less

Onset of runaway nucleation in aerosol reactors

NASA Technical Reports Server (NTRS)

Wu, Jin Jwang; Flagan, Richard C.

1987-01-01

The onset of homogeneous nucleation of new particles from the products of gas phase chemical reactions was explored using an aerosol reactor in which seed particles of silicon were grown by silane pyrolysis. The transition from seed growth by cluster deposition to catastrophic nucleation was extremely abrupt, with as little as a 17 percent change in the reactant concentration leading to an increase in the concentration of measurable particles of four orders of magnitude. From the structure of the particles grown near this transition, it is apparent that much of the growth occurs by the accumulation of clusters on the growing seed particles. The time scale for cluster diffusion indicates, however, that the clusters responsible for growth must be much smaller than the apparent fine structure of the product particles.

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

Kennedy, R.P.; Kincaid, R.H.; Short, S.A.

This report presents the results of part of a two-task study on the engineering characterization of earthquake ground motion for nuclear power plant design. Task I of the study, which is presented in NUREG/CR-3805, Vol. 1, developed a basis for selecting design response spectra taking into account the characteristics of free-field ground motion found to be significant in causing structural damage. Task II incorporates additional considerations of effects of spatial variations of ground motions and soil-structure interaction on foundation motions and structural response. The results of Task II are presented in four parts: (1) effects of ground motion characteristics onmore » structural response of a typical PWR reactor building with localized nonlinearities and soil-structure interaction effects; (2) empirical data on spatial variations of earthquake ground motion; (3) soil-structure interaction effects on structural response; and (4) summary of conclusions and recommendations based on Tasks I and II studies. This report presents the results of the first part of Task II. The results of the other parts will be presented in NUREG/CR-3805, Vols. 3 to 5.« less

Response surface method in geotechnical/structural analysis, phase 1

NASA Astrophysics Data System (ADS)

Wong, F. S.

1981-02-01

In the response surface approach, an approximating function is fit to a long running computer code based on a limited number of code calculations. The approximating function, called the response surface, is then used to replace the code in subsequent repetitive computations required in a statistical analysis. The procedure of the response surface development and feasibility of the method are shown using a sample problem in slop stability which is based on data from centrifuge experiments of model soil slopes and involves five random soil parameters. It is shown that a response surface can be constructed based on as few as four code calculations and that the response surface is computationally extremely efficient compared to the code calculation. Potential applications of this research include probabilistic analysis of dynamic, complex, nonlinear soil/structure systems such as slope stability, liquefaction, and nuclear reactor safety.

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.

FFTF Passive Safety Test Data for Benchmarks for New LMR Designs

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

Wootan, David W.; Casella, Andrew M.

Liquid Metal Reactors (LMRs) continue to be considered as an attractive concept for advanced reactor design. Software packages such as SASSYS are being used to im-prove new LMR designs and operating characteristics. Significant cost and safety im-provements can be realized in advanced liquid metal reactor designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associ-ated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. The FFTF passive safety testing pro-gram was developed to examine howmore » specific design elements influenced dynamic re-activity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results to reactors of current interest. The U.S. Department of En-ergy, Office of Nuclear Energy Advanced Reactor Technology program is in the pro-cess of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Benchmarks based on empirical data gathered during operation of the Fast Flux Test Facility (FFTF) as well as design documents and post-irradiation examination will aid in the validation of these software packages and the models and calculations they produce. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs« less

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.

Iron oxide nanoparticle layer templated by polydopamine spheres: a novel scaffold toward hollow-mesoporous magnetic nanoreactors

NASA Astrophysics Data System (ADS)

Huang, Liang; Ao, Lijiao; Xie, Xiaobin; Gao, Guanhui; Foda, Mohamed F.; Su, Wu

2014-12-01

Superparamagnetic iron oxide nanoparticle layers with high packing density and controlled thickness were in situ deposited on metal-affinity organic templates (polydopamine spheres), via one-pot thermal decomposition. The as synthesized hybrid structure served as a facile nano-scaffold toward hollow-mesoporous magnetic carriers, through surfactant-assisted silica encapsulation and its subsequent calcination. Confined but accessible gold nanoparticles were successfully incorporated into these carriers to form a recyclable catalyst, showing quick magnetic response and a large surface area (642.5 m2 g-1). Current nano-reactors exhibit excellent catalytic performance and high stability in reduction of 4-nitrophenol, together with convenient magnetic separability and good reusability. The integration of compact iron oxide nanoparticle layers with programmable polydopamine templates paves the way to fabricate magnetic-response hollow structures, with high permeability and multi-functionality.Superparamagnetic iron oxide nanoparticle layers with high packing density and controlled thickness were in situ deposited on metal-affinity organic templates (polydopamine spheres), via one-pot thermal decomposition. The as synthesized hybrid structure served as a facile nano-scaffold toward hollow-mesoporous magnetic carriers, through surfactant-assisted silica encapsulation and its subsequent calcination. Confined but accessible gold nanoparticles were successfully incorporated into these carriers to form a recyclable catalyst, showing quick magnetic response and a large surface area (642.5 m2 g-1). Current nano-reactors exhibit excellent catalytic performance and high stability in reduction of 4-nitrophenol, together with convenient magnetic separability and good reusability. The integration of compact iron oxide nanoparticle layers with programmable polydopamine templates paves the way to fabricate magnetic-response hollow structures, with high permeability and multi-functionality. Electronic supplementary information (ESI) available: Fig. S1-S5. See DOI: 10.1039/c4nr05931j

Disturbance Frequency Determines Morphology and Community Development in Multi-Species Biofilm at the Landscape Scale

PubMed Central

Milferstedt, Kim; Santa-Catalina, Gaëlle; Godon, Jean-Jacques; Escudié, Renaud; Bernet, Nicolas

2013-01-01

Many natural and engineered biofilm systems periodically face disturbances. Here we present how the recovery time of a biofilm between disturbances (expressed as disturbance frequency) shapes the development of morphology and community structure in a multi-species biofilm at the landscape scale. It was hypothesized that a high disturbance frequency favors the development of a stable adapted biofilm system while a low disturbance frequency promotes a dynamic biofilm response. Biofilms were grown in laboratory-scale reactors over a period of 55-70 days and exposed to the biocide monochloramine at two frequencies: daily or weekly pulse injections. One untreated reactor served as control. Biofilm morphology and community structure were followed on comparably large biofilm areas at the landscape scale using automated image analysis (spatial gray level dependence matrices) and community fingerprinting (single-strand conformation polymorphisms). We demonstrated that a weekly disturbed biofilm developed a resilient morphology and community structure. Immediately after the disturbance, the biofilm simplified but recovered its initial complex morphology and community structure between two biocide pulses. In the daily treated reactor, one organism largely dominated a morphologically simple and stable biofilm. Disturbances primarily affected the abundance distribution of already present bacterial taxa but did not promote growth of previously undetected organisms. Our work indicates that disturbances can be used as lever to engineer biofilms by maintaining a biofilm between two developmental states. PMID:24303024

Nuclear reactor fuel containment safety structure

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

Rosewell, M.P.

A nuclear reactor fuel containment safety structure is disclosed and is shown to include an atomic reactor fuel shield with a fuel containment chamber and exhaust passage means, and a deactivating containment base attached beneath the fuel reactor shield and having exhaust passages, manifold, and fluxing and control material and vessels. 1 claim, 8 figures.

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.; Morton, T. J.

2006-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG testing will utilize a higher fidelity point kinetics model to control core power transients, and reactivity feedback will be based on localized feedback coefficients and several independent temperature measurements taken within the core block. This paper presents preliminary test results and discusses the methodology that will be implemented in follow-on DDG testing and the additional instrumentation required to implement high fidelity dynamic testing.

Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

NASA Astrophysics Data System (ADS)

Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

2016-01-01

Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

Seismic isolation of small modular reactors using metamaterials

NASA Astrophysics Data System (ADS)

Witarto, Witarto; Wang, S. J.; Yang, C. Y.; Nie, Xin; Mo, Y. L.; Chang, K. C.; Tang, Yu; Kassawara, Robert

2018-04-01

Adaptation of metamaterials at micro- to nanometer scales to metastructures at much larger scales offers a new alternative for seismic isolation systems. These new isolation systems, known as periodic foundations, function both as a structural foundation to support gravitational weight of the superstructure and also as a seismic isolator to isolate the superstructure from incoming seismic waves. Here we describe the application of periodic foundations for the seismic protection of nuclear power plants, in particular small modular reactors (SMR). For this purpose, a large-scale shake table test on a one-dimensional (1D) periodic foundation supporting an SMR building model was conducted. The 1D periodic foundation was designed and fabricated using reinforced concrete and synthetic rubber (polyurethane) materials. The 1D periodic foundation structural system was tested under various input waves, which include white noise, stepped sine and seismic waves in the horizontal and vertical directions as well as in the torsional mode. The shake table test results show that the 1D periodic foundation can reduce the acceleration response (transmissibility) of the SMR building up to 90%. In addition, the periodic foundation-isolated structure also exhibited smaller displacement than the non-isolated SMR building. This study indicates that the challenge faced in developing metastructures can be overcome and the periodic foundations can be applied to isolating vibration response of engineering structures.

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

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.

Application of Simulated Reactivity Feedback in Nonnuclear Testing of a Direct-Drive Gas-Cooled Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Webster, K. L.

2007-01-01

Nonnuclear testing can be a valuable tool in the development of an in-space nuclear power or propulsion system. In a nonnuclear test facility, electric heaters are used to simulate heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response and response characteristics, and assess potential design improvements with a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE 100a heat pipe cooled, electrically heated reactor and heat exchanger hardware. This Technical Memorandum discusses the status of the planned dynamic test methodology for implementation in the direct-drive gas-cooled reactor testing and assesses the additional instrumentation needed to implement high-fidelity dynamic testing.

Test Results from a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

NASA Technical Reports Server (NTRS)

Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.; Godfroy, Thomas J.

2010-01-01

Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

From Reactor to Rheology in LDPE Modeling

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

Read, Daniel J.; Das, Chinmay; Auhl, Dietmar

2008-07-07

In recent years the association between molecular structure and linear rheology has been established and well-understood through the tube concept and its extensions for well-characterized materials (e.g. McLeish, Adv. Phys. 2002). However, for industrial branched polymeric material at processing conditions this piece of information is missing. A large number of phenomenological models have been developed to describe the nonlinear response of polymers. But none of these models takes into account the underlying molecular structure, leading to a fitting procedure with arbitrary fitting parameters. The goal of applied molecular rheology is a predictive scheme that runs in its entirety from themore » molecular structure from the reactor to the non-linear rheology of the resin. In our approach, we use a model for the industrial reactor to explicitly generate the molecular structure ensemble of LDPE's, (Tobita, J. Polym. Sci. B 2001), which are consistent with the analytical information. We calculate the linear rheology of the LDPE ensemble with the use of a tube model for branched polymers (Das et al., J. Rheol. 2006). We then, separate the contribution of the stress decay to a large number of pompom modes (McLeish et al., J. Rheol. 1998 and Inkson et al., J. Rheol. 1999) with the stretch time and the priority variables corresponding to the actual ensemble of molecules involved. This multimode pompom model allows us to predict the nonlinear properties without any fitting parameter. We present and analyze our results in comparison with experimental data on industrial materials.« less

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.

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.

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.

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.

A high resolution pneumatic stepping actuator for harsh reactor environments

NASA Astrophysics Data System (ADS)

Tippetts, Thomas B.; Evans, Paul S.; Riffle, George K.

1993-01-01

A reactivity control actuator for a high-power density nuclear propulsion reactor must be installed in close proximity to the reactor core. The energy input from radiation to the actuator structure could exceed hundreds of W/cc unless low-cross section, low-absorptivity materials are chosen. Also, for post-test handling and subsequent storage, materials should not be used that are activated into long half-life isotopes. Pneumatic actuators can be constructed from various reactor-compatible materials, but conventional pneumatic piston actuators generally lack the stiffness required for high resolution reactivity control unless electrical position sensors and compensated electronic control systems are used. To overcome these limitations, a pneumatic actuator is under development that positions an output shaft in response to a series of pneumatic pulses, comprising a pneumatic analog of an electrical stepping motor. The pneumatic pulses are generated remotely, beyond the strong radiation environment, and transmitted to the actuator through tubing. The mechanically simple actuator uses a nutating gear harmonic drive to convert motion of small pistons directly to high-resolution angular motion of the output shaft. The digital nature of this actuator is suitable for various reactor control algorithms but is especially compatible with the three bean salad algorithm discussed by Ball et al. (1991).

The Use Of High Speed Photography In Reactor Safety Studies At The Atomic Energy Establishment, Winfrith

NASA Astrophysics Data System (ADS)

Maddison, R. J.

1985-02-01

The investigation of certain areas of nuclear reactor safety involves the study of high speed phenomena with timescales ranging from microseconds to a few hundreds of milliseconds. Examples which have been extensively studied at Winfrith are firstly, the thermal interaction of molten fuel and reactor coolant which can generate high pressures on the 100 msec timescale, and which involves phenomena such as vapour film collapse which takes place on the microsecond timescale. Secondly, there is the response of reactor structures to such pressures, and finally there is the response of structural materials such as metals and concrete to the impulsive loading arising from the impact of heavy, high velocity missiles. A wide range of experimental techniques is used in these studies, many of which have been developed specially for this type of work which ranges from small laboratory scale to large field scale experiments. There are two important features which characterise many of these experiments:- i) a long period of meticulous preparation of very heavily instrumented, short duration experiments and; ii) the destructive nature of the experiments. Various forms of High Speed photography are included in the inventory of experimental techniques. These include the use of single and double exposure, short duration, spark photography; the use of an Image Convertor Camera (IMACON 790); and a number of rotating prism cine cameras. High Speed Photography is used both in a primary experimental role in the studies, and in a supportive role for other instrumentation. Because of the sometimes violent nature of these experiments, cameras are often heavily protected and operated remotely; lighting systems are sometimes destroyed. This has led to the development of unconventional techniques for camera operation and subject lighting. This paper will describe some of the experiments and the way in which High Speed Photography has been applied as an essential experimental tool. It will be illustrated with cine film taken during the experiments.

Eukaryotic Community Shift in Response to Organic Loading Rate of an Aerobic Trickling Filter (Down-Flow Hanging Sponge Reactor) Treating Domestic Sewage.

PubMed

Miyaoka, Yuma; Hatamoto, Masashi; Yamaguchi, Takashi; Syutsubo, Kazuaki

2017-05-01

In this study, changes in eukaryotic community structure and water quality were investigated in an aerobic trickling filter (down-flow hanging sponge, DHS) treating domestic sewage under different organic loading rates (OLRs). The OLR clearly influenced both sponge pore water quality and relative flagellates and ciliates (free-swimming, carnivorous, crawling, and stalked protozoa) abundances in the retained sludge. Immediately after the OLR was increased from 1.05 to 1.97 kg chemical oxygen demand (COD) m -3  day -1 , COD and NH 4 + -N treatment efficiencies both deteriorated, and relative flagellates and ciliates abundances then increased from 2-8 % to 51-65 % total cells in the middle-bottom part of the DHS reactor. In a continuous operation at a stable OLR (2.01 kg COD m -3  day -1 ), effluent water quality improved, and relative flagellates and ciliates abundances decreased to 15-46 % total cells in the middle-bottom part of the DHS reactor. This result may indicate that flagellates and ciliates preferentially graze on dispersed bacteria, thus, stabilizing effluent water quality. Additionally, to investigate eukaryotic community structure, clone libraries based on the 18S ribosomal ribonucleic acid (rRNA) gene of the retained sludge were constructed. The predominant group was Nucletmycea phylotypes, representing approximately 29-56 % total clones. Furthermore, a large proportion of the clones had <97 % sequence identity in the NCBI database. This result indicates that phylogenetically unknown eukaryotes were present in the DHS reactor. These results provide insights into eukaryotic community shift in the DHS reactor treating domestic sewage.

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.

Integrated ‘omics analysis for studying the microbial community response to a pH perturbation of a cellulose-degrading bioreactor culture

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

Boaro, Amy A.; Kim, Young-Mo; Konopka, Allan E.

2015-01-05

Propionate accumulation is a common indicator of process imbalances in anaerobic bioreactor systems. The accumulation of propionate can occur due to low retention rates, hydrogen accumulation, or mechanical changes affecting the proximity between propionate oxidizers and partner species, thereby preventing necessary electron transfer. Few studies, however, have observed the changes in microbial community structure during propionate accumulation. We used 454 pyrosequencing of 16S rDNA to evaluate the community membership during propionate accumulations in replicate bioreactors with rumen based cultures. Half of the culture volume from a parent reactor was transferred to a sterile “daughter” reactor, and both systems were runmore » identically. Both reactors experienced a propionate accumulation after roughly 10 days, with the propionate accumulation being less pronounced in the parent reactor as compared to the daughter reactor. Non-metric multidimensional scaling (NMDS) was used to determine clustering patterns of the samples, and correlative methods were used to determine which OTUs were significantly associated with the movements of samples along the NMDS axes. The presence of Saccharofermentans characterized the position of early samples, whereas the presence of Ruminococcus and Succiniclasticum were more indicative of the positions of later samples. Hydrogen accumulation and low sequence counts indicated low methanogen activity. Although both reactor systems were closed to microbial inputs due to the sterilization of influent media, we recorded significant increases in reactor diversity over time. This suggests that changes in the abundances of dominant community members may affect the sequencing of rare taxa within samples.« less

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.

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

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.

Core reactivity estimation in space reactors using recurrent dynamic networks

NASA Technical Reports Server (NTRS)

Parlos, Alexander G.; Tsai, Wei K.

1991-01-01

A recurrent multilayer perceptron network topology is used in the identification of nonlinear dynamic systems from only the input/output measurements. The identification is performed in the discrete time domain, with the learning algorithm being a modified form of the back propagation (BP) rule. The recurrent dynamic network (RDN) developed is applied for the total core reactivity prediction of a spacecraft reactor from only neutronic power level measurements. Results indicate that the RDN can reproduce the nonlinear response of the reactor while keeping the number of nodes roughly equal to the relative order of the system. As accuracy requirements are increased, the number of required nodes also increases, however, the order of the RDN necessary to obtain such results is still in the same order of magnitude as the order of the mathematical model of the system. It is believed that use of the recurrent MLP structure with a variety of different learning algorithms may prove useful in utilizing artificial neural networks for recognition, classification, and prediction of dynamic systems.

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.

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

Collecting and recirculating condensate in a nuclear reactor containment

DOEpatents

Schultz, Terry L.

1993-01-01

An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank.

Collecting and recirculating condensate in a nuclear reactor containment

DOEpatents

Schultz, T.L.

1993-10-19

An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank. 3 figures.

Cross-flow electrochemical reactor cells, cross-flow reactors, and use of cross-flow reactors for oxidation reactions

DOEpatents

Balachandran, Uthamalingam; Poeppel, Roger B.; Kleefisch, Mark S.; Kobylinski, Thaddeus P.; Udovich, Carl A.

1994-01-01

This invention discloses cross-flow electrochemical reactor cells containing oxygen permeable materials which have both electron conductivity and oxygen ion conductivity, cross-flow reactors, and electrochemical processes using cross-flow reactor cells having oxygen permeable monolithic cores to control and facilitate transport of oxygen from an oxygen-containing gas stream to oxidation reactions of organic compounds in another gas stream. These cross-flow electrochemical reactors comprise a hollow ceramic blade positioned across a gas stream flow or a stack of crossed hollow ceramic blades containing a channel or channels for flow of gas streams. Each channel has at least one channel wall disposed between a channel and a portion of an outer surface of the ceramic blade, or a common wall with adjacent blades in a stack comprising a gas-impervious mixed metal oxide material of a perovskite structure having electron conductivity and oxygen ion conductivity. The invention includes reactors comprising first and second zones seprated by gas-impervious mixed metal oxide material material having electron conductivity and oxygen ion conductivity. Prefered gas-impervious materials comprise at least one mixed metal oxide having a perovskite structure or perovskite-like structure. The invention includes, also, oxidation processes controlled by using these electrochemical reactors, and these reactions do not require an external source of electrical potential or any external electric circuit for oxidation to proceed.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2009-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2010-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

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.

The Shock and Vibration Digest, Volume 4, Number 6, June 1972.

DTIC Science & Technology

1972-06-01

mooring system. THE NUCLEAR SHIP MUTSU AD-736117 Nara, S. (Ishikawajima-Harima Heavy Industries Co., Ltd., Tokyo, Japan) 72-1031 Nippon Zosen Gakkaishi...sideration In the nuclear ship Mutsu , and response, space shuttles collision-resistant structures are provided on both sides of the reactor room and...are three methods to absorb is utilized for stability derivative estimation. collision energy effectively, and in Mutsu , side The damping in the

Nuclear reactor melt-retention structure to mitigate direct containment heating

DOEpatents

Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

1991-01-01

A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

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

Light Water Reactor Sustainability Program: Survey of Models for Concrete Degradation

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

Spencer, Benjamin W.; Huang, Hai

Concrete is widely used in the construction of nuclear facilities because of its structural strength and its ability to shield radiation. The use of concrete in nuclear facilities for containment and shielding of radiation and radioactive materials has made its performance crucial for the safe operation of the facility. As such, when life extension is considered for nuclear power plants, it is critical to have predictive tools to address concerns related to aging processes of concrete structures and the capacity of structures subjected to age-related degradation. The goal of this report is to review and document the main aging mechanismsmore » of concern for concrete structures in nuclear power plants (NPPs) and the models used in simulations of concrete aging and structural response of degraded concrete structures. This is in preparation for future work to develop and apply models for aging processes and response of aged NPP concrete structures in the Grizzly code. To that end, this report also provides recommendations for developing more robust predictive models for aging effects of performance of concrete.« less

Sulfamethoxazole and COD increase abundance of sulfonamide resistance genes and change bacterial community structures within sequencing batch reactors.

PubMed

Guo, Xueping; Pang, Weihai; Dou, Chunling; Yin, Daqiang

2017-05-01

The abundant microbial community in biological treatment processes in wastewater treatment plants (WWTPs) may potentially enhance the horizontal gene transfer of antibiotic resistance genes with the presence of antibiotics. A lab-scale sequencing batch reactor was designed to investigate response of sulfonamide resistance genes (sulI, sulII) and bacterial communities to various concentrations of sulfamethoxazole (SMX) and chemical oxygen demand (COD) of wastewater. The SMX concentrations (0.001 mg/L, 0.1 mg/L and 10 mg/L) decreased with treatment time and higher SMX level was more difficult to remove. The presence of SMX also significantly reduced the removal efficiency of ammonia nitrogen, affecting the normal function of WWTPs. All three concentrations of SMX raised both sulI and sulII genes with higher concentrations exhibiting greater increases. The abundance of sul genes was positive correlated with treatment time and followed the second-order reaction kinetic model. Interestingly, these two genes have rather similar activity. SulI and sulII gene abundance also performed similar response to COD. Simpson index and Shannon-Weiner index did not show changes in the microbial community diversity. However, the 16S rRNA gene cloning and sequencing results showed the bacterial community structures varied during different stages. The results demonstrated that influent antibiotics into WWTPs may facilitate selection of ARGs and affect the wastewater conventional treatment as well as the bacteria community structures. Copyright © 2017 Elsevier Ltd. All rights reserved.

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

Telling the public about risks

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

Ahearne, J.F.

1990-09-01

The Three Mile Island and Chernobyl reactor accidents are used along with several non-nuclear hazards and risks to illustrate that, indeed, communication is a two-way process. The interplay of bureaucratic structures and the U.S. legal system is one cause of the problem. Dispersion of responsibility when two or more agencies are involved is another problem. Even when a single agency is responsible, the available level citizen participation may be murky. Communications must be honest; credibility is strengthened by honesty and lost by lying. But between these two ends of the spectrum be persuasion, manipulation, and deceit. 3 refs.

SAFETY SYSTEM FOR CONTROL ROD

DOEpatents

Paget, J.A.

1963-05-14

A structure for monitoring the structural continuity of a control rod foi a neutron reactor is presented. A electric conductor readily breakable under mechanical stress is fastened along the length of the control rod at a plurality of positions and forms a closed circuit with remote electrical components responsive to an open circuit. A portion of the conductor between the control rod and said components is helically wound to allow free and normally unrestricted movement of the segment of conductor secured to the control rod relative to the remote components. Any break in the circuit is indicative of control rod breakage. (AEC)

U.S. Nuclear Weapons: Changes in Policy and Force Structure

DTIC Science & Technology

2008-01-23

Pinellas Plant , in Clearwater, FL; and the Pantex Plant near Amarillo, TX. These facilities were also operated by industrial contractors. Finally, the...These included the Rocky Flats Plant , outside Denver, CO; the Kansas City Plant , near Kansas City, MO; the Mound Plant , near Dayton OH; the...In 1988, DOE closed the nuclear reactors at Hanford and Savannah River, in response to safety concerns. The Rocky Flats Plant , which produced the

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

Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

DOEpatents

McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

1994-05-03

The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

DOEpatents

McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

1994-01-01

The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

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

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.

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.

Correlating Fast Fluence to dpa in Atypical Locations

NASA Astrophysics Data System (ADS)

Drury, Thomas H.

2016-02-01

Damage to a nuclear reactor's materials by high-energy neutrons causes changes in the ductility and fracture toughness of the materials. The reactor vessel and its associated piping's ability to withstand stress without brittle fracture are paramount to safety. Theoretically, the material damage is directly related to the displacements per atom (dpa) via the residual defects from induced displacements. However in practice, the material damage is based on a correlation to the high-energy (E > 1.0 MeV) neutron fluence. While the correlated approach is applicable when the material in question has experienced the same neutron spectrum as test specimens which were the basis of the correlation, this approach is not generically acceptable. Using Monte Carlo and discrete ordinates transport codes, the energy dependent neutron flux is determined throughout the reactor structures and the reactor vessel. Results from the models provide the dpa response in addition to the high-energy neutron flux. Ratios of dpa to fast fluence are calculated throughout the models. The comparisons show a constant ratio in the areas of historical concern and thus the validity of the correlated approach to these areas. In regions above and below the fuel however, the flux spectrum has changed significantly. The correlated relationship of material damage to fluence is not valid in these regions without adjustment. An adjustment mechanism is proposed.

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

Improved Nuclear Reactor and Shield Mass Model for Space Applications

NASA Technical Reports Server (NTRS)

Robb, Kevin

2004-01-01

New technologies are being developed to explore the distant reaches of the solar system. Beyond Mars, solar energy is inadequate to power advanced scientific instruments. One technology that can meet the energy requirements is the space nuclear reactor. The nuclear reactor is used as a heat source for which a heat-to-electricity conversion system is needed. Examples of such conversion systems are the Brayton, Rankine, and Stirling cycles. Since launch cost is proportional to the amount of mass to lift, mass is always a concern in designing spacecraft. Estimations of system masses are an important part in determining the feasibility of a design. I worked under Michael Barrett in the Thermal Energy Conversion Branch of the Power & Electric Propulsion Division. An in-house Closed Cycle Engine Program (CCEP) is used for the design and performance analysis of closed-Brayton-cycle energy conversion systems for space applications. This program also calculates the system mass including the heat source. CCEP uses the subroutine RSMASS, which has been updated to RSMASS-D, to estimate the mass of the reactor. RSMASS was developed in 1986 at Sandia National Laboratories to quickly estimate the mass of multi-megawatt nuclear reactors for space applications. In response to an emphasis for lower power reactors, RSMASS-D was developed in 1997 and is based off of the SP-100 liquid metal cooled reactor. The subroutine calculates the mass of reactor components such as the safety systems, instrumentation and control, radiation shield, structure, reflector, and core. The major improvements in RSMASS-D are that it uses higher fidelity calculations, is easier to use, and automatically optimizes the systems mass. RSMASS-D is accurate within 15% of actual data while RSMASS is only accurate within 50%. My goal this summer was to learn FORTRAN 77 programming language and update the CCEP program with the RSMASS-D model.

EMERGENCY SHUTDOWN FOR NUCLEAR REACTORS

DOEpatents

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

1963-12-24

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

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).

Energy-technological complex with reactor for torrefaction

NASA Astrophysics Data System (ADS)

Kuzmina, J. S.; Director, L. B.; Zaichenko, V. M.

2016-11-01

To eliminate shortcomings of raw plant materials pelletizing process with thermal treatment (low-temperature pyrolysis or torrefaction) can be applied. This paper presents a mathematical model of energy-technological complex (ETC) for combined production of heat, electricity and solid biofuels torrefied pellets. According to the structure the mathematical model consists of mathematical models of main units of ETC and the relationships between them and equations of energy and material balances. The equations describe exhaust gas straining action through a porous medium formed by pellets. Decomposition rate of biomass was calculated by using the gross-reaction diagram, which is responsible for the disintegration of raw material. A mathematical model has been tested according to bench experiments on one reactor module. From nomographs, designed for a particular configuration of ETC it is possible to determine the basic characteristics of torrefied pellets (rate of weight loss, heating value and heat content) specifying only two parameters (temperature and torrefaction time). It is shown that the addition of reactor for torrefaction to gas piston engine can improve the energy efficiency of power plant.

Responses to diagnostic tests for bovine tuberculosis in dairy and non-dairy cattle naturally exposed to Mycobacterium bovis in Great Britain.

PubMed

Downs, S H; Broughan, J M; Goodchild, A V; Upton, P A; Durr, P A

2016-10-01

Field surveillance of British cattle using the single intradermal comparative cervical tuberculin (SICCT) test shows a higher incidence rate of bovine tuberculosis (bTB) in dairy compared to beef herds, but a lower probability of post-mortem examination confirmed (PMC) Mycobacterium bovis infection in dairy herds. A cross-sectional study was conducted to compare animal level differences in bTB detection between dairy and non-dairy cattle in Great Britain. During the period from 2002 to 2005, 200 (41% dairy) reactors in the SICCT test (standard interpretation) were randomly selected, and 200 in-contact cattle (43% dairy) were purposively selected from bTB-infected herds. Interferon (IFN)-γ responses in blood to bovine and avian purified protein derivative (PPD), and early secretory antigen target 6 kDa and culture filtrate protein 10 (ESAT-6/CFP10), were measured. The post-mortem examination included gross pathological examination, mycobacterial culture and histopathology. The proportions of cattle positive to ESAT6/CFP10 were 26% (95% confidence interval, CI, 15-39%) in dairy reactors and 62% (95% CI 51-72%) in non-dairy reactors (P <0.001). PMC risk was 34% (95% CI 24-45%) in dairy reactors and 69% (95% CI 60-78%) in non-dairy reactors (P <0.001). The odds ratio for PMC risk in dairy reactors compared to non-dairy reactors, after controlling for bTB prevalence, herd size and SICCT test response, was 0.27 (95% CI 0.14-0.53; P <0.001). In surveillance data, adjusted animal level PMC risks were lower for dairy reactors than for beef reactors aged >2 years (P <0.001). Crown Copyright © 2016. Published by Elsevier Ltd. All rights reserved.

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).

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

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.

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.

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…

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.

Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

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

Maloy, Stuart Andrew

In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

Reactor vessel support system. [LMFBR

DOEpatents

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

1980-05-09

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

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.

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.

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.

Optical Sensors for Monitoring Gamma and Neutron Radiation

NASA Technical Reports Server (NTRS)

Boyd, Clark D.

2011-01-01

For safety and efficiency, nuclear reactors must be carefully monitored to provide feedback that enables the fission rate to be held at a constant target level via adjustments in the position of neutron-absorbing rods and moderating coolant flow rates. For automated reactor control, the monitoring system should provide calibrated analog or digital output. The sensors must survive and produce reliable output with minimal drift for at least one to two years, for replacement only during refueling. Small sensor size is preferred to enable more sensors to be placed in the core for more detailed characterization of the local fission rate and fuel consumption, since local deviations from the norm tend to amplify themselves. Currently, reactors are monitored by local power range meters (LPRMs) based on the neutron flux or gamma thermometers based on the gamma flux. LPRMs tend to be bulky, while gamma thermometers are subject to unwanted drift. Both electronic reactor sensors are plagued by electrical noise induced by ionizing radiation near the reactor core. A fiber optic sensor system was developed that is capable of tracking thermal neutron fluence and gamma flux in order to monitor nuclear reactor fission rates. The system provides near-real-time feedback from small- profile probes that are not sensitive to electromagnetic noise. The key novel feature is the practical design of fiber optic radiation sensors. The use of an actinoid element to monitor neutron flux in fiber optic EFPI (extrinsic Fabry-Perot interferometric) sensors is a new use of material. The materials and structure used in the sensor construction can be adjusted to result in a sensor that is sensitive to just thermal, gamma, or neutron stimulus, or any combination of the three. The tested design showed low sensitivity to thermal and gamma stimuli and high sensitivity to neutrons, with a fast response time.

The close relation between Lactococcus and Methanosaeta is a keystone for stable methane production from molasses wastewater in a UASB reactor.

PubMed

Kim, Tae Gwan; Yun, Jeonghee; Cho, Kyung-Suk

2015-10-01

The up-flow anaerobic sludge blanket (UASB) reactor is a promising method for the treatment of high-strength industrial wastewaters due to advantage of its high treatment capacity and settleable suspended biomass retention. Molasses wastewater as a sugar-rich waste is one of the most valuable raw material for bioenergy production due to its high organic strength and bioavailability. Interpretation for complex interactions of microbial community structures and operational parameters can help to establish stable biogas production. RNA-based approach for biogas production systems is recommended for analysis of functionally active community members which are significantly underestimated. In this study, methane production and active microbial community were characterized in an UASB reactor using molasses wastewater as feedstock. The UASB reactor achieved a stable process performance at an organic loading rate of 1.7~13.8-g chemical oxygen demand (COD,·L(-1) day(-1); 87-95 % COD removal efficiencies), and the maximum methane production rate was 4.01 L-CH4·at 13.8 g-COD L(-1) day(-1). Lactococcus and Methanosaeta were comprised up to 84 and 80 % of the active bacterial and archaeal communities, respectively. Network analysis of reactor performance and microbial community revealed that Lactococcus and Methanosaeta were network hub nodes and positively correlated each other. In addition, they were positively correlated with methane production and organic loading rate, and they shared the other microbial hub nodes as neighbors. The results indicate that the close association between Lactococcus and Methanosaeta is responsible for the stable production of methane in the UASB reactor using molasses wastewater.

NSR&D Program Fiscal Year (FY) 2015 Call for Proposals Mitigation of Seismic Risk at Nuclear Facilities using Seismic Isolation

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

Coleman, Justin

2015-02-01

Seismic isolation (SI) has the potential to drastically reduce seismic response of structures, systems, or components (SSCs) and therefore the risk associated with large seismic events (large seismic event could be defined as the design basis earthquake (DBE) and/or the beyond design basis earthquake (BDBE) depending on the site location). This would correspond to a potential increase in nuclear safety by minimizing the structural response and thus minimizing the risk of material release during large seismic events that have uncertainty associated with their magnitude and frequency. The national consensus standard America Society of Civil Engineers (ASCE) Standard 4, Seismic Analysismore » of Safety Related Nuclear Structures recently incorporated language and commentary for seismically isolating a large light water reactor or similar large nuclear structure. Some potential benefits of SI are: 1) substantially decoupling the SSC from the earthquake hazard thus decreasing risk of material release during large earthquakes, 2) cost savings for the facility and/or equipment, and 3) applicability to both nuclear (current and next generation) and high hazard non-nuclear facilities. Issue: To date no one has evaluated how the benefit of seismic risk reduction reduces cost to construct a nuclear facility. Objective: Use seismic probabilistic risk assessment (SPRA) to evaluate the reduction in seismic risk and estimate potential cost savings of seismic isolation of a generic nuclear facility. This project would leverage ongoing Idaho National Laboratory (INL) activities that are developing advanced (SPRA) methods using Nonlinear Soil-Structure Interaction (NLSSI) analysis. Technical Approach: The proposed study is intended to obtain an estimate on the reduction in seismic risk and construction cost that might be achieved by seismically isolating a nuclear facility. The nuclear facility is a representative pressurized water reactor building nuclear power plant (NPP) structure. Figure 1: Project activities The study will consider a representative NPP reinforced concrete reactor building and representative plant safety system. This study will leverage existing research and development (R&D) activities at INL. Figure 1 shows the proposed study steps with the steps in blue representing activities already funded at INL and the steps in purple the activities that would be funded under this proposal. The following results will be documented: 1) Comparison of seismic risk for the non-seismically isolated (non-SI) and seismically isolated (SI) NPP, and 2) an estimate of construction cost savings when implementing SI at the site of the generic NPP.« less

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

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

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

Gusev, S. I.; Karpov, V. N.; Kiselev, A. N.

2009-09-15

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

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

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)

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.

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.

FLOW SYSTEM FOR REACTOR

DOEpatents

Zinn, W.H.

1963-06-11

A reactor is designed with means for terminating the reaction when returning coolant is below a predetermined temperature. Coolant flowing from the reactor passes through a heat exchanger to a lower reservoir, and then circulates between the lower reservoir and an upper reservoir before being returned to the reactor. Means responsive to the temperature of the coolant in the return conduit terminate the chain reaction when the temperature reaches a predetermined minimum value. (AEC)

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.

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.

Fission-powered in-core thermoacoustic sensor

DOE PAGES

Garrett, Steven L.; Smith, James A.; Smith, Robert W. M.; ...

2016-04-07

A thermoacoustic engine is operated within the core of a nuclear reactor to acoustically telemeter coolant temperature (frequency-encoded) and reactor power level (amplitude-encoded) outside the reactor, thus providing the values of these important parameters without external electrical power or wiring. We present data from two hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. Furthermore, these signals have been detected even in the presence of substantial background noise generated by the reactor's fluid pumps.

Fission-powered in-core thermoacoustic sensor

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

Garrett, Steven L.; Smith, James A.; Smith, Robert W. M.

2016-04-04

A thermoacoustic engine is operated within the core of a nuclear reactor to acoustically telemeter coolant temperature (frequency-encoded) and reactor power level (amplitude-encoded) outside the reactor, thus providing the values of these important parameters without external electrical power or wiring. We present data from two hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These signals have been detected even in the presence of substantial background noise generated by the reactor's fluid pumps.

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

Effects of Predation by Protists on Prokaryotic Community Function, Structure, and Diversity in Anaerobic Granular Sludge.

PubMed

Hirakata, Yuga; Oshiki, Mamoru; Kuroda, Kyohei; Hatamoto, Masashi; Kubota, Kengo; Yamaguchi, Takashi; Harada, Hideki; Araki, Nobuo

2016-09-29

Predation by protists is top-down pressure that regulates prokaryotic abundance, community function, structure, and diversity in natural and artificial ecosystems. Although the effects of predation by protists have been studied in aerobic ecosystems, they are poorly understood in anoxic environments. We herein studied the influence of predation by Metopus and Caenomorpha ciliates-ciliates frequently found in anoxic ecosystems-on prokaryotic community function, structure, and diversity. Metopus and Caenomorpha ciliates were cocultivated with prokaryotic assemblages (i.e., anaerobic granular sludge) in an up-flow anaerobic sludge blanket (UASB) reactor for 171 d. Predation by these ciliates increased the methanogenic activities of granular sludge, which constituted 155% of those found in a UASB reactor without the ciliates (i.e., control reactor). Sequencing of 16S rRNA gene amplicons using Illumina MiSeq revealed that the prokaryotic community in the UASB reactor with the ciliates was more diverse than that in the control reactor; 2,885-3,190 and 2,387-2,426 operational taxonomic units (>97% sequence similarities), respectively. The effects of predation by protists in anaerobic engineered systems have mostly been overlooked, and our results show that the influence of predation by protists needs to be examined and considered in the future for a better understanding of prokaryotic community structure and function.

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1974-01-01

The paper summarizes the results obtained in an exploratory evaluation of ceramics for automobile thermal reactors. Candidate ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance, lasting 1100 hours in engine dynamometer tests and for more than 38,600 kilimeters (24,000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1974-01-01

The results obtained in an exploratory evaluation of ceramics for automobile thermal reactors are summarized. Candidate ceramic materials were evaluated in several reactor designs by using both engine-dynamometer and vehicle road tests. Silicon carbide contained in a corrugated-metal support structure exhibited the best performance, lasting 1100 hr in engine-dynamometer tests and more than 38,600 km (24000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as those containing silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

Multi-layered mode structure of locked-tearing-modes after unlocking

NASA Astrophysics Data System (ADS)

Okabayashi, Michio; Logan, N.; Tobias, B.; Wang, Z.; Budny, B.; Nazikian, R.; Strait, E.; La Haye, R.; Paz-Soldan, C. J.; Ferraro, N.; Shiraki, D.; Hanson, J.; Zanca, P.; Paccagnella, R.

2015-11-01

Prevention of m/n=2/1 tearing modes (TM) by electro-magnetic torque injection has been successful in DIII-D and RFX-mod where plasma conditions and plasma shape are completely different. Understanding the internal structure in the post-unlocked phase is a pre-requisite to its application to reactor relevant plasmas such as in ITER. Ti and toroidal rotation perturbations show there exist several radially different TM layers. However, the phase shift between the applied field and the plasma response is rather small from plasma edge to the q ~3 domain, indicating that a kink-like response prevails. The biggest threat for sustaining an unlocked 2/1 mode is sudden distortion of the rotational profile due to the internal mode reconnection. Possible TM layer structure will be discussed with numerical MHD codes and TRANSP. This work is supported in part by the US Department of Energy under DE-AC02-09CH11466, DE-FG02-99ER54531, DE-SC0003913, and DE-FC02-04ER54698.

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

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

Chang, S.J.

1997-05-01

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

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.

Nuclear reactor

DOEpatents

Wade, Elman E.

1979-01-01

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

N Reactor Deactivation Program Plan. Revision 4

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

Walsh, J.L.

1993-12-01

This N Reactor Deactivation Program Plan is structured to provide the basic methodology required to place N Reactor and supporting facilities {center_dot} in a radiologically and environmentally safe condition such that they can be decommissioned at a later date. Deactivation will be in accordance with facility transfer criteria specified in Department of Energy (DOE) and Westinghouse Hanford Company (WHC) guidance. Transition activities primarily involve shutdown and isolation of operational systems and buildings, radiological/hazardous waste cleanup, N Fuel Basin stabilization and environmental stabilization of the facilities. The N Reactor Deactivation Program covers the period FY 1992 through FY 1997. The directivemore » to cease N Reactor preservation and prepare for decommissioning was issued by DOE to WHC on September 20, 1991. The work year and budget data supporting the Work Breakdown Structure in this document are found in the Activity Data Sheets (ADS) and the Environmental Restoration Program Baseline, that are prepared annually.« less

NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

DOEpatents

Moore, W.T.

1958-09-01

This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

Radioactive Waste Management and Nuclear Facility Decommissioning Progress in Iraq - 13216

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

Al-Musawi, Fouad; Shamsaldin, Emad S.; Jasim, Hadi

2013-07-01

Management of Iraq's radioactive wastes and decommissioning of Iraq's former nuclear facilities are the responsibility of Iraq's Ministry of Science and Technology (MoST). The majority of Iraq's former nuclear facilities are in the Al-Tuwaitha Nuclear Research Center located a few kilometers from the edge of Baghdad. These facilities include bombed and partially destroyed research reactors, a fuel fabrication facility and radioisotope production facilities. Within these facilities are large numbers of silos, approximately 30 process or waste storage tanks and thousands of drums of uncharacterised radioactive waste. There are also former nuclear facilities/sites that are outside of Al-Tuwaitha and these includemore » the former uranium processing and waste storage facility at Jesira, the dump site near Adaya, the former centrifuge facility at Rashdiya and the former enrichment plant at Tarmiya. In 2005, Iraq lacked the infrastructure needed to decommission its nuclear facilities and manage its radioactive wastes. The lack of infrastructure included: (1) the lack of an organization responsible for decommissioning and radioactive waste management, (2) the lack of a storage facility for radioactive wastes, (3) the lack of professionals with experience in decommissioning and modern waste management practices, (4) the lack of laws and regulations governing decommissioning or radioactive waste management, (5) ongoing security concerns, and (6) limited availability of electricity and internet. Since its creation eight years ago, the MoST has worked with the international community and developed an organizational structure, trained staff, and made great progress in managing radioactive wastes and decommissioning Iraq's former nuclear facilities. This progress has been made, despite the very difficult implementing conditions in Iraq. Within MoST, the Radioactive Waste Treatment and Management Directorate (RWTMD) is responsible for waste management and the Iraqi Decommissioning Directorate (IDD) is responsible for decommissioning activities. The IDD and the RWTMD work together on decommissioning projects. The IDD has developed plans and has completed decommissioning of the GeoPilot Facility in Baghdad and the Active Metallurgical Testing Laboratory (LAMA) in Al-Tuwaitha. Given this experience, the IDD has initiated work on more dangerous facilities. Plans are being developed to characterize, decontaminate and decommission the Tamuz II Research Reactor. The Tammuz Reactor was destroyed by an Israeli air-strike in 1981 and the Tammuz II Reactor was destroyed during the First Gulf War in 1991. In addition to being responsible for managing the decommissioning wastes, the RWTMD is responsible for more than 950 disused sealed radioactive sources, contaminated debris from the first Gulf War and (approximately 900 tons) of naturally-occurring radioactive materials wastes from oil production in Iraq. The RWTMD has trained staff, rehabilitated the Building 39 Radioactive Waste Storage building, rehabilitated portions of the French-built Radioactive Waste Treatment Station, organized and secured thousands of drums of radioactive waste organized and secured the stores of disused sealed radioactive sources. Currently, the IDD and the RWTMD are finalizing plans for the decommissioning of the Tammuz II Research Reactor. (authors)« less

Early Program Development

NASA Image and Video Library

1963-01-01

This artist's concept from 1963 shows a proposed NERVA (Nuclear Engine for Rocket Vehicle Application) incorporating the NRX-A1, the first NERVA-type cold flow reactor. The NERVA engine, based on Kiwi nuclear reactor technology, was intended to power a RIFT (Reactor-In-Flight-Test) nuclear stage, for which Marshall Space Flight Center had development responsibility.

Measurement of neutron spectra in the experimental reactor LR-0

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

Prenosil, Vaclav; Mravec, Filip; Veskrna, Martin

2015-07-01

The measurement of fast neutron fluxes is important in many areas of nuclear technology. It affects the stability of the reactor structural components, performance of fuel, and also the fuel manner. The experiments performed at the LR-0 reactor were in the past focused on the measurement of neutron field far from the core, in reactor pressure vessel simulator or in biological shielding simulator. In the present the measurement in closer regions to core became more important, especially measurements in structural components like reactor baffle. This importance increases with both reactor power increase and also long term operation. Other important taskmore » is an increasing need for the measurement close to the fuel. The spectra near the fuel are aimed due to the planned measurements with the FLIBE salt, in FHR / MSR research, where one of the task is the measurement of the neutron spectra in it. In both types of experiments there is strong demand for high working count rate. The high count rate is caused mainly by high gamma background and by high fluxes. The fluxes in core or in its vicinity are relatively high to ensure safe reactor operation. This request is met in the digital spectroscopic apparatus. All experiments were realized in the LR-0 reactor. It is an extremely flexible light water zero-power research reactor, operated by the Research Center Rez (Czech Republic). (authors)« less

A laboratory investigation of interactions between denitrifying anaerobic methane oxidation (DAMO) and anammox processes in anoxic environments

PubMed Central

Hu, Shihu; Zeng, Raymond J.; Haroon, Mohamed F.; Keller, Jurg; Lant, Paul A.; Tyson, Gene W.; Yuan, Zhiguo

2015-01-01

This study investigates interactions between recently identified denitrifying anaerobic methane oxidation (DAMO) and anaerobic ammonium oxidation (anammox) processes in controlled anoxic laboratory reactors. Two reactors were seeded with the same inocula containing DAMO organisms Candidatus Methanoperedens nitroreducens and Candidatus Methylomirabilis oxyfera, and anammox organism Candidatus Kuenenia stuttgartiensis. Both were fed with ammonium and methane, but one was also fed with nitrate and the other with nitrite, providing anoxic environments with different electron acceptors. After steady state reached in several months, the DAMO process became solely/primarily responsible for nitrate reduction while the anammox process became solely responsible for nitrite reduction in both reactors. 16S rRNA gene amplicon sequencing showed that the nitrate-driven DAMO organism M. nitroreducens dominated both the nitrate-fed (~70%) and the nitrite-fed (~26%) reactors, while the nitrite-driven DAMO organism M. oxyfera disappeared in both communities. The elimination of M. oxyfera from both reactors was likely the results of this organism being outcompeted by anammox bacteria for nitrite. K. stuttgartiensis was detected at relatively low levels (1–3%) in both reactors. PMID:25732131

Activation characteristics of candidate structural materials for a near-term Indian fusion reactor and the impact of their impurities on design considerations

NASA Astrophysics Data System (ADS)

H, L. SWAMI; C, DANANI; A, K. SHAW

2018-06-01

Activation analyses play a vital role in nuclear reactor design. Activation analyses, along with nuclear analyses, provide important information for nuclear safety and maintenance strategies. Activation analyses also help in the selection of materials for a nuclear reactor, by providing the radioactivity and dose rate levels after irradiation. This information is important to help define maintenance activity for different parts of the reactor, and to plan decommissioning and radioactive waste disposal strategies. The study of activation analyses of candidate structural materials for near-term fusion reactors or ITER is equally essential, due to the presence of a high-energy neutron environment which makes decisive demands on material selection. This study comprises two parts; in the first part the activation characteristics, in a fusion radiation environment, of several elements which are widely present in structural materials, are studied. It reveals that the presence of a few specific elements in a material can diminish its feasibility for use in the nuclear environment. The second part of the study concentrates on activation analyses of candidate structural materials for near-term fusion reactors and their comparison in fusion radiation conditions. The structural materials selected for this study, i.e. India-specific Reduced Activation Ferritic‑Martensitic steel (IN-RAFMS), P91-grade steel, stainless steel 316LN ITER-grade (SS-316LN-IG), stainless steel 316L and stainless steel 304, are candidates for use in ITER either in vessel components or test blanket systems. Tungsten is also included in this study because of its use for ITER plasma-facing components. The study is carried out using the reference parameters of the ITER fusion reactor. The activation characteristics of the materials are assessed considering the irradiation at an ITER equatorial port. The presence of elements like Nb, Mo, Co and Ta in a structural material enhance the activity level as well as the dose level, which has an impact on design considerations. IN-RAFMS was shown to be a more effective low-activation material than SS-316LN-IG.

Nuclear reactor apparatus

DOEpatents

Wade, Elman E.

1978-01-01

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

NEUTRONIC REACTOR

DOEpatents

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

1961-11-21

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

Nuclear reactor shield including magnesium oxide

DOEpatents

Rouse, Carl A.; Simnad, Massoud T.

1981-01-01

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

Irradiation response of commercial, high-Tc superconducting tapes: Electromagnetic transport properties

DOE PAGES

Gapud, A. A.; Greenwood, N. T.; Alexander, J. A.; ...

2015-07-01

Effects of low dose irradiation on the electrical transport current properties of commercially available high-temperature superconducting, coated-conductor tapes were investigated, in view of potential applications in the irradiative environment of fusion reactors. Three different tapes, each with unique as-grown flux-pinning structures, were irradiated with Au and Ni ions at energies that provide a range of damage effects, with accumulated damage levels near that expected for conductors in a fusion reactor environment. Measurements using transport current determined the pre- and post-irradiation resistivity, critical current density, and pinning force density, yielding critical temperatures, irreversibility lines, and inferred vortex creep rates. Results showmore » that at the irradiation damage levels tested, any detriment to as-grown pre-irradiation properties is modest; indeed in one case already-superior pinning forces are enhanced, leading to higher critical currents.« less

Light water reactor lower head failure analysis

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

Rempe, J.L.; Chavez, S.A.; Thinnes, G.L.

1993-10-01

This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broadermore » range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.« less

Light Water Reactor Sustainability Program: Risk-Informed Safety Margins Characterization (RISMC) Pathway Technical Program Plan

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

Smith, Curtis; Rabiti, Cristian; Martineau, Richard

Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). As the current Light Water Reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of Systems, Structures, and Components (SSCs) degradations or failures that initiate safety-significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degreemore » of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated, primarily based on “engineering judgment.”« less

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

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.

Detection of neutrinos, antineutrinos, and neutrino-like particles

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

Fischbach, Ephraim

An apparatus for detecting the presence of a nuclear reactor by the detection of antineutrinos from the reactor can include a radioactive sample having a measurable nuclear activity level and a decay rate capable of changing in response to the presence of antineutrinos, and a detector associated with the radioactive sample. The detector is responsive to at least one of a particle or radiation formed by decay of the radioactive sample. A processor associated with the detector can correlate rate of decay of the radioactive sample to a flux of the antineutrinos to detect the reactor.

METHOD FOR SENSING DEGREE OF FLUIDIZATION IN FLUIDIZED BED

DOEpatents

Levey, R.P. Jr.; Fowler, A.H.

1961-12-12

A method is given for detecting, indicating, and controlling the degree of fluidization in a fluid-bed reactor into which powdered material is fed. The method comprises admitting of gas into the reactor, inserting a springsupported rod into the powder bed of the reactor, exciting the rod to vibrate at its resonant frequency, deriving a signal responsive to the amplitude of vibi-ation of the rod and spring, the signal being directiy proportional to the rate of flow of the gas through the reactor, displaying the signal to provide an indication of the degree of fluidization within the reactor, and controlling the rate of gas flow into the reactor until said signal stabilizes at a constant value to provide substantially complete fluidization within the reactor. (AEC)

Jacking mechanism for upper internals structure of a liquid metal nuclear reactor

DOEpatents

Gillett, James E.; Wineman, Arthur L.

1984-01-01

A jacking mechanism for raising the upper internals structure of a liquid metal nuclear reactor which jacking mechanism uses a system of gears and drive shafts to transmit force from a single motor to four mechanically synchronized ball jacks to raise and lower support columns which support the upper internals structure. The support columns have a pin structure which rides up and down in a slot in a housing fixed to the reactor head. The pin has two locking plates which can be rotated around the pin to bring bolt holes through the locking plates into alignment with a set of bolt holes in the housing, there being a set of such housing bolt holes corresponding to both a raised and a lowered position of the support column. When the locking plate is so aligned, a surface of the locking plate mates with a surface in the housing such that the support column is then supported by the locking plate and not by the ball jacks. Since the locking plates are to be installed and bolted to the housing during periods of reactor operation, the ball jacks need not be sized to react the large forces which occur or potentially could occur on the upper internals structure of the reactor during operation. The locking plates react these loads. The ball jacks, used only during refueling, can be smaller, which enable conventionally available equipment to fulfill the precision requirements for the task within available space.

Development of Inspection and Repair Technology for Heat Exchanger Tubes in Fast Breeder Reactors

DTIC Science & Technology

2009-06-01

Technology for Heat Exchanger Tubes in Fast Breeder Reactors Akihiko NISHIMURA *1 , Takahisa SHOBU, Kiyoshi OKA, Toshihiko YAMAGUCHI, Yukihiro SHIMADA...fast breeder reactors (FBRs). It comprises a laser processing head combined with an eddy current testing unit. Ultrashort laser pulse ablation is used...be applied in the main- tenance of large structures such as nuclear reactors and chemical factories [1]. Internal access to a blanket cooling pipe

Exploratory evaluation of ceramics for automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1972-01-01

An exploratory evaluation of ceramics for automobile thermal reactors was conducted. Potential ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance lasting over 800 hours in engine dynamometer tests and over 15,000 miles (24,200 km) of vehicle road tests. Reactors containing glass-ceramic components did not perform as well as silicon carbide. But the glass-ceramics still offer good potential for reactor use. The results of this study are considered to be a reasonable demonstration of the potential use of ceramics in thermal reactors.

Nuclear energy center site survey reactor plant considerations

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

Harty, H.

The Energy Reorganization Act of 1974 required the Nuclear Regulatory Commission (NRC) to make a nuclear energy center site survey (NECSS). Background information for the NECSS report was developed in a series of tasks which include: socioeconomic inpacts; environmental impact (reactor facilities); emergency response capability (reactor facilities); aging of nuclear energy centers; and dry cooled nuclear energy centers.

CONTROL SYSTEM

DOEpatents

Shannon, R.H.; Williamson, H.E.

1962-10-30

A boiling water type nuclear reactor power system having improved means of control is described. These means include provisions for either heating the coolant-moderator prior to entry into the reactor or shunting the coolantmoderator around the heating means in response to the demand from the heat engine. These provisions are in addition to means for withdrawing the control rods from the reactor. (AEC)

BWR Anticipated Transients Without Scram Leading to Instability

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

Cheng L. Y.; Baek J.; Cuadra, A.

2013-11-10

Anticipated transients without scram (ATWS) in aboiling water reactor (BWR) were simulated in order to understand reactor response and determine the effectiveness of automatic and operator actions to mitigate this beyond-design-basis accident. The events of interest herein are initiated by a turbine trip when the reactor is operating in the expanded operating domainMELLLA+ [maximum extended load line limit plus]. In these events the reactor may initially be at up to 120% of the original licensed thermal power (OLTP) and at flow rates as low as 80% of rated.For these (and similar) ATWS events the concern isthat when the reactor powermore » decreases in response to a dual recirculation pump trip, the core will become unstable and large amplitude oscillations will begin. The occurrence of these power oscillations, if left unmitigated, may result in fuel damage, and the amplitude of the poweroscillations may hamper the effectiveness of the injection of dissolved neutron absorber through the standby liquid control system (SLCS).« less

Researcher Poses with a Nuclear Rocket Model

NASA Image and Video Library

1961-11-21

A researcher at the NASA Lewis Research Center with slide ruler poses with models of the earth and a nuclear-propelled rocket. The Nuclear Engine for Rocket Vehicle Applications (NERVA) was a joint NASA and Atomic Energy Commission (AEC) endeavor to develop a nuclear-powered rocket for both long-range missions to Mars and as a possible upper-stage for the Apollo Program. The early portion of the program consisted of basic reactor and fuel system research. This was followed by a series of Kiwi reactors built to test nuclear rocket principles in a non-flying nuclear engine. The next phase, NERVA, would create an entire flyable engine. The AEC was responsible for designing the nuclear reactor and overall engine. NASA Lewis was responsible for developing the liquid-hydrogen fuel system. The nuclear rocket model in this photograph includes a reactor at the far right with a hydrogen propellant tank and large radiator below. The payload or crew would be at the far left, distanced from the reactor.

Magnetically-induced forces on a ferromagnetic HT-9 first wall/blanket module

NASA Astrophysics Data System (ADS)

Lechtenberg, T. A.; Dahms, C. F.; Attaya, H.

1984-05-01

A model of the Starfire commercial tokamak reactor was used as the basis for calculating magnetic loads induced on typical fusion reactor first wall components fabricated of ferromagnetic material. The component analyzed was the first wall/blanket module because this structure experiences the greatest neutron fluence level and is the component for which the low swelling ferromagnetic Sandvik alloy, HT-9, may have the greatest benefit. The magnitudes of the magnetic body forces calculated were consistent with analyses performed on structures within other types of reactors. The loads generated within the module structure by the magnetic forces were found to be of the same order of magnitude as those arising from other sources such as pressure differential, dead weight, temperature distribution. Only small structural design modifications would be required if the magnetic alloy, Sandvik HT-9 were utilized.

APPARATUS FOR CONTROL OF A BOILING REACTOR RESPONSIVE TO STEAM DEMAND

DOEpatents

Treshow, M.

1963-07-23

A method of controlling a fuel-rod-in-tube-type boilingwater reactor having nozzles at the point of water entry into the tube is described. Water is pumped into the nozzles by an auxiliary pump operated by steam from an interstage position of the associated turbine, so that the pumping speed is responsive to turbine demand. (AEC)

Methane and hydrogen sulfide emissions in UASB reactors treating domestic wastewater.

PubMed

Souza, C L; Chernicharo, C A L; Melo, G C B

2012-01-01

The release of CH(4) and H(2)S in UASB reactors was evaluated with the aim to quantify the emissions from the liquid surfaces (three-phase separator and settler compartment) and also from the reactor's discharge hydraulic structures. The studies were carried out in two pilot- (360 L) and one demo-scale (14 m(3)) UASB reactors treating domestic wastewater. As expected, the release rates were much higher across the gas/liquid interfaces of the three-phase separators (5.4-9.7 kg CH(4) m(-2) d(-1) and 23.0-35.8 g S m(-2) d(-1)) as compared with the quiescent settler surfaces (11.0-17.8 g CH(4) m(-2) d(-1) and 0.21 to 0.37 g S m(-2) d(-1)). The decrease of dissolved methane and dissolved hydrogen sulfide was very large in the discharging hydraulic structures very close to the reactor (>60 and >80%, respectively), largely due to the loss to the atmosphere, indicating that the concentration of these compounds will probably fall to values close to zero in the near downstream structures. The emission factors due to the release of dissolved methane in the discharge structure amounted to around 0.040 g CH(4) g COD(infl)(-1) and 0.060 g CH(4) g COD(rem)(-1), representing around 60% of the methane collected in the three-phase separator.

Structural materials for Gen-IV nuclear reactors: Challenges and opportunities

NASA Astrophysics Data System (ADS)

Murty, K. L.; Charit, I.

2008-12-01

Generation-IV reactor design concepts envisioned thus far cater toward 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-IV 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 environment, 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 paper presents a summary of various Gen-IV reactor concepts, with emphasis on the structural materials issues depending on the specific application areas. This paper also discusses the challenges involved in using the existing materials under both service and off-normal conditions. Tasks become increasingly complex due to the operation of various fundamental phenomena like radiation-induced segregation, radiation-enhanced diffusion, precipitation, interactions between impurity elements and radiation-produced defects, swelling, helium generation and so forth. Further, high temperature capability (e.g. creep properties) of these materials is a critical, performance-limiting factor. It is demonstrated that novel alloy and microstructural design approaches coupled with new materials processing and fabrication techniques may mitigate the challenges, and the optimum system performance may be achieved under much demanding conditions.

10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL ...

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

10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL CROSS SECTION. Giffals & Vallet, Inc., L. Rosetti, Associated Architects and Engineers, Detroit, Michigan; and U.S. Army Engineer Division, New England Corps of Engineers, Boston, Massachusetts. Drawing Number 35-84-04. (Original: AMTL Engineering Division, Watertown). - Watertown Arsenal, Building No. 100, Wooley Avenue, Watertown, Middlesex County, MA

Passive cooling system for nuclear reactor containment structure

DOEpatents

Gou, Perng-Fei; Wade, Gentry E.

1989-01-01

A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

Extension of the TRANSURANUS burnup model to heavy water reactor conditions

NASA Astrophysics Data System (ADS)

Lassmann, K.; Walker, C. T.; van de Laar, J.

1998-06-01

The extension of the light water reactor burnup equations of the TRANSURANUS code to heavy water reactor conditions is described. Existing models for the fission of 235U and the buildup of plutonium in a heavy water reactor are evaluated. In order to overcome the limitations of the frequently used RADAR model at high burnup, a new model is presented. After verification against data for the radial distributions of Xe, Cs, Nd and Pu from electron probe microanalysis, the model is used to analyse the formation of the high burnup structure in a heavy water reactor. The new model allows the analysis of light water reactor fuel rod designs at high burnup in the OECD Halden Heavy Water Reactor.

MEANS FOR CONTROLLING A NUCLEAR REACTOR

DOEpatents

Wilson, V.C.; Overbeck, W.P.; Slotin, L.; Froman, D.K.

1957-12-17

This patent relates to nuclear reactors of the type using a solid neutron absorbing material as a means for controlling the reproduction ratio of the system and thereby the power output. Elongated rods of neutron absorbing material, such as boron steel for example, are adapted to be inserted and removed from the core of tae reactor by electronic motors and suitable drive means. The motors and drive means are controlled by means responsive to the neutron density, such as ionization chambers. The control system is designed to be responsive also to the rate of change in neutron density to automatically maintain the total power output at a substantially constant predetermined value. A safety rod means responsive to neutron density is also provided for keeping the power output below a predetermined maximum value at all times.

Safe Affordable Fission Engine-(SAFE-) 100a Heat Exchanger Thermal and Structural Analysis

NASA Technical Reports Server (NTRS)

Steeve, B. E.

2005-01-01

A potential fission power system for in-space missions is a heat pipe-cooled reactor coupled to a Brayton cycle. In this system, a heat exchanger (HX) transfers the heat of the reactor core to the Brayton gas. The Safe Affordable Fission Engine- (SAFE-) 100a is a test program designed to thermally and hydraulically simulate a 95 Btu/s prototypic heat pipe-cooled reactor using electrical resistance heaters on the ground. This Technical Memorandum documents the thermal and structural assessment of the HX used in the SAFE-100a program.

Psychosocial effects of restarting a TMI reactor

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

Not Available

1990-01-01

ORNL is studying human responses to hazardous environmental phenomena. This study attempts to understand the human behavior associated with the restart of TMI-1 Reactor after a nuclear event occurred at TMI-2.

HEDL FACILITIES CATALOG 400 AREA

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

MAYANCSIK BA

1987-03-01

The purpose of this project is to provide a sodium-cooled fast flux test reactor designed specifically for irradiation testing of fuels and materials and for long-term testing and evaluation of plant components and systems for the Liquid Metal Reactor (LMR) Program. The FFTF includes the reactor, heat removal equipment and structures, containment, core component handling and examination, instrumentation and control, and utilities and other essential services. The complex array of buildings and equipment are arranged around the Reactor Containment Building.

MCNP-REN - A Monte Carlo Tool for Neutron Detector Design Without Using the Point Model

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

Abhold, M.E.; Baker, M.C.

1999-07-25

The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo N-Particle code (MCNP) was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP - Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program (TAP) predict neutron detector response without using the pointmore » reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of MOX fresh fuel made using the Underwater Coincidence Counter (UWCC) as well as measurements of HEU reactor fuel using the active neutron Research Reactor Fuel Counter (RRFC) are compared with calculations. The method used in MCNP-REN is demonstrated to be fundamentally sound and shown to eliminate the need to use the point model for detector performance predictions.« less

Solvent refined coal (SRC) process. Annual technical progress report, January 1979-December 1979

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

Not Available

1980-11-01

A set of statistically designed experiments was used to study the effects of several important operating variables on coal liquefaction product yield structures. These studies used a Continuous Stirred-Tank Reactor to provide a hydrodynamically well-defined system from which kinetic data could be extracted. An analysis of the data shows that product yield structures can be adequately represented by a correlative model. It was shown that second-order effects (interaction and squared terms) are necessary to provide a good model fit of the data throughout the range studied. Three reports were issued covering the SRC-II database and yields as functions of operatingmore » variables. The results agree well with the generally-held concepts of the SRC reaction process, i.e., liquid phase hydrogenolysis of liquid coal which is time-dependent, thermally activated, catalyzed by recycle ash, and reaction rate-controlled. Four reports were issued summarizing the comprehensive SRC reactor thermal response models and reporting the results of several studies made with the models. Analytical equipment for measuring SRC off-gas composition and simulated distillation of coal liquids and appropriate procedures have been established.« less

Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

NASA Astrophysics Data System (ADS)

Tong, H.; Snow, G. C.; Chu, E. K.; Chang, R. L. S.; Angwin, M. J.; Pessagno, S. L.

1981-09-01

Durable catalytic reactors for advanced gas turbine engines were developed. Objectives were: to evaluate furnace aging as a cost effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1,000 hours of combustion durability, and define a catalytic reactor system with a high probability of successful integration into an automotive gas turbine engine. Fourteen different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel for combustion conditions. Eight additional catalytic reactors were evaluated and one of these was successfully combustion tested on propane fuel. This durability reactor used graded cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

Development of a high-temperature durable catalyst for use in catalytic combustors for advanced automotive gas turbine engines

NASA Technical Reports Server (NTRS)

Tong, H.; Snow, G. C.; Chu, E. K.; Chang, R. L. S.; Angwin, M. J.; Pessagno, S. L.

1981-01-01

Durable catalytic reactors for advanced gas turbine engines were developed. Objectives were: to evaluate furnace aging as a cost effective catalytic reactor screening test, measure reactor degradation as a function of furnace aging, demonstrate 1,000 hours of combustion durability, and define a catalytic reactor system with a high probability of successful integration into an automotive gas turbine engine. Fourteen different catalytic reactor concepts were evaluated, leading to the selection of one for a durability combustion test with diesel fuel for combustion conditions. Eight additional catalytic reactors were evaluated and one of these was successfully combustion tested on propane fuel. This durability reactor used graded cell honeycombs and a combination of noble metal and metal oxide catalysts. The reactor was catalytically active and structurally sound at the end of the durability test.

PBF (PER620) interior. Detail view across top of reactor tank. ...

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

PBF (PER-620) interior. Detail view across top of reactor tank. Camera facing northeast. Ait tubing is cleanup equipment. Note projections from reactor structure above water level in tank. Date: May 2004. INEEL negative no. HD-41-5-1 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2012 CFR

2012-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2013 CFR

2013-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2014 CFR

2014-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

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.

Globally linearized control on diabatic continuous stirred tank reactor: a case study.

PubMed

Jana, Amiya Kumar; Samanta, Amar Nath; Ganguly, Saibal

2005-07-01

This paper focuses on the promise of globally linearized control (GLC) structure in the realm of strongly nonlinear reactor system control. The proposed nonlinear control strategy is comprised of: (i) an input-output linearizing state feedback law (transformer), (ii) a state observer, and (iii) an external linear controller. The synthesis of discrete-time GLC controller for single-input single-output diabatic continuous stirred tank reactor (DCSTR) has been studied first, followed by the synthesis of feedforward/feedback controller for the same reactor having dead time in process as well as in disturbance. Subsequently, the multivariable GLC structure has been designed and then applied on multi-input multi-output DCSTR system. The simulation study shows high quality performance of the derived nonlinear controllers. The better-performed GLC in conjunction with reduced-order observer has been compared with the conventional proportional integral controller on the example reactor and superior performance has been achieved by the proposed GLC control scheme.

Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors

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

William Richins; Stephen Novascone; Cheryl O'Brien

Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors William Richins1, Stephen Novascone1, and Cheryl O’Brien1 1Idaho National Laboratory, US Dept. of Energy, Idaho Falls, Idaho, USA, e-mail: William.Richins@inl.gov The Idaho National Laboratory (INL, USA) and IASMiRT sponsored an international forum Nov 5-6, 2008 in Porvoo, Finland for nuclear industry, academic, and regulatory representatives to identify structural issues in current and future advanced reactor design, especially for extreme conditions and external threats. The purpose of this Topical Workshop was to articulate research, engineering, and regulatory Code development needs. The topics addressed by the Workshop were selectedmore » to address critical industry needs specific to advanced reactor structures that have long lead times and can be the subject of future SMiRT technical sessions. The topics were; 1) structural/materials needs for extreme conditions and external threats in contemporary (Gen. III) and future (Gen. IV and NGNP) advanced reactors and 2) calibrating simulation software and methods that address topic 1 The workshop discussions and research needs identified are presented. The Workshop successfully produced interactive discussion on the two topics resulting in a list of research and technology needs. It is recommended that IASMiRT communicate the results of the discussion to industry and researchers to encourage new ideas and projects. In addition, opportunities exist to retrieve research reports and information that currently exists, and encourage more international cooperation and collaboration. It is recommended that IASMiRT continue with an off-year workshop series on select topics.« less

Microbial community dynamics and biogas production from manure fractions in sludge bed anaerobic digestion.

PubMed

Nordgård, A S R; Bergland, W H; Bakke, R; Vadstein, O; Østgaard, K; Bakke, I

2015-12-01

To elucidate how granular sludge inoculum and particle-rich organic loading affect the structure of the microbial communities and process performance in upflow anaerobic sludge bed (UASB) reactors. We investigated four reactors run on dairy manure filtrate and four on pig manure supernatant for three months achieving similar methane yields. The reactors fed with less particle rich pig manure stabilized faster and had highest capacity. Microbial community dynamics analysed by a PCR/denaturing gradient gel electrophoresis approach showed that influent was a major determinant for the composition of the reactor communities. Comparisons of pre- and non-adapted inoculum in the reactors run on pig manure supernatant showed that the community structure of the nonadapted inoculum adapted in approximately two months. Microbiota variance partitioning analysis revealed that running time, organic loading rate and inoculum together explained 26 and 31% of the variance in bacterial and archaeal communities respectively. The microbial communities of UASBs adapted to the reactor conditions in treatment of particle rich manure fractions, obtaining high capacity, especially on pig manure supernatant. These findings provide relevant insight into the microbial community dynamics in startup and operation of sludge bed reactors for methane production from slurry fractions, a major potential source of biogas. © 2015 The Society for Applied Microbiology.

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

Reactor Design and Decommissioning - An Overview of International Activities in Post Fukushima Era1 - 12396

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

Devgun, Jas S.; Laraia, Michele; Pescatore, Claudio

Accidents at the Fukushima Dai-ichi reactors as a result of the devastating earthquake and tsunami of March 11, 2011 have not only dampened the nuclear renaissance but have also initiated a re-examination of the design and safety features for the existing and planned nuclear reactors. Even though failures of some of the key site features at Fukushima can be attributed to events that in the past would have been considered as beyond the design basis, the industry as well as the regulatory authorities are analyzing what features, especially passive features, should be designed into the new reactor designs to minimizemore » the potential for catastrophic failures. It is also recognized that since the design of the Fukushima BWR reactors which were commissioned in 1971, many advanced safety features are now a part of the newer reactor designs. As the recovery efforts at the Fukushima site are still underway, decisions with respect to the dismantlement and decommissioning of the damaged reactors and structures have not yet been finalized. As it was with Three Mile Island, it could take several decades for dismantlement, decommissioning and clean up, and the project poses especially tough challenges. Near-term assessments have been issued by several organizations, including the IAEA, the USNRC and others. Results of such investigations will lead to additional improvements in system and site design measures including strengthening of the anti-tsunami defenses, more defense-in-depth features in reactor design, and better response planning and preparation involving reactor sites. The question also arises what would the effect be on the decommissioning scene worldwide, and what would the effect be on the new reactors when they are eventually retired and dismantled. This paper provides an overview of the US and international activities related to recovery and decommissioning including the decommissioning features in the reactor design process and examines these from a new perspective in the post Fukushima -accident era. Accidents at the Fukushima Daiichi reactors in the aftermath of the devastating earthquake and tsunami of March 11, 2011 have slowed down the nuclear renaissance world-wide and may have accelerated decommissioning either because some countries have decided to halt or reduce nuclear, or because the new safety requirements may reduce life-time extensions. Even in countries such as the UK and France that favor nuclear energy production existing nuclear sites are more likely to be chosen as sites for future NPPs. Even as the site recovery efforts continue at Fukushima and any decommissioning decisions are farther into the future, the accidents have focused attention on the reactor designs in general and specifically on the Fukushima type BWRs. The regulatory authorities in many countries have initiated a re-examination of the design of the systems, structures and components and considerations of the capability of the station to cope with beyond-design basis events. Enhancements to SSCs and site features for the existing reactors and the reactors that will be built will also impact the decommissioning phase activities. The newer reactor designs of today not only have enhanced safety features but also take into consideration the features that will facilitate future decommissioning. Lessons learned from past management and operation of reactors as well as the lessons from decommissioning are incorporated into the new designs. However, in the post-Fukushima era, the emphasis on beyond-design-basis capability may lead to significant changes in SSCs, which eventually will also have impact on the decommissioning phase. Additionally, where some countries decide to phase out the nuclear power, many reactors may enter the decommissioning phase in the coming decade. While the formal updating and expanding of existing guidance documents for accident cleanup and decommissioning would benefit by waiting until the Fukushima project has progressed sufficiently for that experience to be reliably interpreted, the development of structured on-line sharing of information and especially the creation of an on-line compendium of methods, tools, and techniques by which damaged fuel and other unique situations have been addressed can be addressed sooner and maintained as new problems and solutions arise and are resolved. The IAEA's new 'WEB 2.0 tool' CONNECT is expected to play a significant role in this and related information-sharing activities. The trend in some countries such as the United States has been to re-license the existing reactors for additional twenty years, beyond the original design life. Given the advances in technology over the past four decades, and considering that the newer designs incorporate significant improvements in safety systems, it may not be economical or technically feasible to retrofit enhancements into some of the older reactors. In such cases, the reactors may be retired from service and decommissioned. Overall, the energy demand in the world continues to rise, with sharp increases in the Asian countries, and nuclear power's role in the world's energy supply is expected to continue. Events at Fukushima have led to a re-examination on many fronts, including reactor design and regulatory requirements. Further changes may occur in these areas in the post-Fukushima era. These changes in turn will also impact the world-wide decommissioning scene and the decommissioning phase of the future reactors. (authors)« less

NUCLEAR REACTOR

DOEpatents

Grebe, J.J.

1961-01-24

A core structure for neutronic reactors adapted for the propulsion of aircraft and rockets is offered. The core is designed for cooling by gaseous media, and comprises a plurality of hollow tapered tubular segments of a porous moderating material impregniated with fissionable fuel nested about a common axis. Alternate ends of the segments are joined. In operation a coolant gas passes through the porous structure and is heated.

Natural circulating passive cooling system for nuclear reactor containment structure

DOEpatents

Gou, Perng-Fei; Wade, Gentry E.

1990-01-01

A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

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.

Handbook explaining the fundamentals of nuclear and atomic physics

NASA Technical Reports Server (NTRS)

Hanlen, D. F.; Morse, W. J.

1969-01-01

Indoctrination document presents nuclear, reactor, and atomic physics in an easy, straightforward manner. The entire subject of nuclear physics including atomic structure ionization, isotopes, radioactivity, and reactor dynamics is discussed.

Double differential light charged particle emission cross sections for some structural fusion materials

NASA Astrophysics Data System (ADS)

Sarpün, Ismail Hakki; n, Abdullah Aydı; Tel, Eyyup

2017-09-01

In fusion reactors, neutron induced radioactivity strongly depends on the irradiated material. So, a proper selection of structural materials will have been limited the radioactive inventory in a fusion reactor. First-wall and blanket components have high radioactivity concentration due to being the most flux-exposed structures. The main objective of fusion structural material research is the development and selection of materials for reactor components with good thermo-mechanical and physical properties, coupled with low-activation characteristics. Double differential light charged particle emission cross section, which is a fundamental data to determine nuclear heating and material damages in structural fusion material research, for some elements target nuclei have been calculated by the TALYS 1.8 nuclear reaction code at 14-15 MeV neutron incident energy and compared with available experimental data in EXFOR library. Direct, compound and pre-equilibrium reaction contribution have been theoretically calculated and dominant contribution have been determined for each emission of proton, deuteron and alpha particle.

Rotatable multifunctional load frames for neutron diffractometers at FRM II—design, specifications and applications

NASA Astrophysics Data System (ADS)

Hoelzel, M.; Gan, W. M.; Hofmann, M.; Randau, C.; Seidl, G.; Jüttner, Ph.; Schmahl, W. W.

2013-05-01

Novel tensile rigs have been designed and manufactured at the research reactor Heinz Maier-Leibnitz (FRM II, Garching near Munich). Besides tensile and compressive stress, also torsion can be applied. The unique Eulerian cradle type design (ω, χ, and φ axis) allows orienting the stress axis with respect to the scattering vector. Applications of these tensile rigs at our neutron diffractometers enable various investigations of structural changes under mechanical load, e.g. crystallographic texture evolution, stress-induced phase transformations or lattice expansion, and the anisotropy of mechanical response.

Reactor Pressure Vessel Integrity Assessments with the Grizzly Aging Simulation Code

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

Spencer, Benjamin; Backman, Marie; Hoffman, William

Grizzly is a simulation tool being developed at Idaho National Laboratory (INL) as part of the US Department of Energy’s Light Water Reactor Sustainability program to provide improved safety assessments of systems, components, and structures in nuclear power plants subjected to age-related degradation. Its goal is to provide an improved scientific basis for decisions surrounding license renewal, which would permit operation of commercial nuclear power plants beyond 60 years. Grizzly is based on INL’s MOOSE framework, which enables multiphysics simulations in a parallel computing environment. It will address a wide variety of aging issues in nuclear power plant systems, components,more » and structures, modelling both the aging processes and the ability of age-degraded components to perform safely. The reactor pressure vessel (RPV) was chosen as the initial application for Grizzly. Grizzly solves tightly coupled equations of heat conduction and solid mechanics to simulate the global response of the RPV to accident conditions, and uses submodels to represent regions with pre-existing flaws. Domain integrals are used to calculate stress intensity factors on those flaws. A physically based empirical model is used to evaluate material embrittlement, and is used to evaluate whether crack growth would occur. Grizzly can represent the RPV in 2D or 3D, allowing it to evaluate effects that require higher dimensionality models to capture. Work is underway to use lower length scale models of material evolution to inform engineering models of embrittlement. This paper demonstrates an application of Grizzly to RPV failure assessment, and summarizes on-going work.« less

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.

The synthesis of cadmium sulfide nanoplatelets using a novel continuous flow sonochemical reactor

DOE PAGES

Palanisamy, Barath; Paul, Brian; Chang, Chih -hung

2015-01-21

A continuous flow sonochemical reactor was developed capable of producing metastable cadmium sulfide (CdS) nanoplatelets with thicknesses at or below 10 nm. The continuous flow sonochemical reactor included the passive in-line micromixing of reagents prior to sonochemical reaction. Synthesis results were compared with those from reactors involving batch conventional heating and batch ultrasound-induced heating. The continuous sonochemical synthesis was found to result in high aspect ratio hexagonal platelets of CdS possessing cubic crystal structures with thicknesses well below 10 nm. The unique shape and crystal structure of the nanoplatelets are suggestive of high localized temperatures within the sonochemical process. Asmore » a result, the particle size uniformity and product throughput are much higher for the continuous sonochemical process in comparison to the batch sonochemical process and conventional synthesis processes.« less

Preliminary Comparison of Radioactive Waste Disposal Cost for Fusion and Fission Reactors

NASA Astrophysics Data System (ADS)

Seki, Yasushi; Aoki, Isao; Yamano, Naoki; Tabara, Takashi

1997-09-01

The environmental and economic impact of radioactive waste (radwaste) generated from fusion power reactors using five types of structural materials and a fission reactor has been evaluated and compared. Possible radwaste disposal scenario of fusion radwaste in Japan is considered. The exposure doses were evaluated for the skyshine of gamma-ray during the disposal operation, groundwater migration scenario during the institutional control period of 300 years and future site use scenario after the institutional period. The radwaste generated from a typical light water fission reactor was evaluated using the same methodology as for the fusion reactors. It is found that radwaste from the fusion reactors using F82H and SiC/SiC composites without impurities could be disposed by the shallow land disposal presently applied to the low level waste in Japan. The disposal cost of radwaste from five fusion power reactors and a typical light water reactor were roughly evaluated and compared.

Microchannel Reactors for ISRU Applications

NASA Astrophysics Data System (ADS)

Carranza, Susana; Makel, Darby B.; Blizman, Brandon; Ward, Benjamin J.

2005-02-01

Affordable planning and execution of prolonged manned space missions depend upon the utilization of local resources and the waste products which are formed in manned spacecraft and surface bases. Successful in-situ resources utilization (ISRU) will require component technologies which provide optimal size, weight, volume, and power efficiency. Microchannel reactors enable the efficient chemical processing of in situ resources. The reactors can be designed for the processes that generate the most benefit for each mission. For instance, propellants (methane) can be produced from carbon dioxide from the Mars atmosphere using the Sabatier reaction and ethylene can be produced from the partial oxidation of methane. A system that synthesizes ethylene could be the precursor for systems to synthesize ethanol and polyethylene. Ethanol can be used as a nutrient for Astrobiology experiments, as well as the production of nutrients for human crew (e.g. sugars). Polyethylene can be used in the construction of habitats, tools, and replacement parts. This paper will present recent developments in miniature chemical reactors using advanced Micro Electro Mechanical Systems (MEMS) and microchannel technology to support ISRU of Mars and lunar missions. Among other applications, the technology has been demonstrated for the Sabatier process and for the partial oxidation of methane. Microchannel reactors were developed based on ceramic substrates as well as metal substrates. In both types of reactors, multiple layers coated with catalytic material are bonded, forming a monolithic structure. Such reactors are readily scalable with the incorporation of extra layers. In addition, this reactor structure minimizes pressure drop and catalyst settling, which are common problems in conventional packed bed reactors.

Precipitation behavior of AlxCoCrFeNi high entropy alloys under ion irradiation

NASA Astrophysics Data System (ADS)

Yang, Tengfei; Xia, Songqin; Liu, Shi; Wang, Chenxu; Liu, Shaoshuai; Fang, Yuan; Zhang, Yong; Xue, Jianming; Yan, Sha; Wang, Yugang

2016-08-01

Materials performance is central to the satisfactory operation of current and future nuclear energy systems due to the severe irradiation environment in reactors. Searching for structural materials with excellent irradiation tolerance is crucial for developing the next generation nuclear reactors. Here, we report the irradiation responses of a novel multi-component alloy system, high entropy alloy (HEA) AlxCoCrFeNi (x = 0.1, 0.75 and 1.5), focusing on their precipitation behavior. It is found that the single phase system, Al0.1CoCrFeNi, exhibits a great phase stability against ion irradiation. No precipitate is observed even at the highest fluence. In contrast, numerous coherent precipitates are present in both multi-phase HEAs. Based on the irradiation-induced/enhanced precipitation theory, the excellent structural stability against precipitation of Al0.1CoCrFeNi is attributed to the high configurational entropy and low atomic diffusion, which reduces the thermodynamic driving force and kinetically restrains the formation of precipitate, respectively. For the multiphase HEAs, the phase separations and formation of ordered phases reduce the system configurational entropy, resulting in the similar precipitation behavior with corresponding binary or ternary conventional alloys. This study demonstrates the structural stability of single-phase HEAs under irradiation and provides important implications for searching for HEAs with higher irradiation tolerance.

Archaeal community composition affects the function of anaerobic co-digesters in response to organic overload

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

Lerm, S.; Kleyboecker, A.; Miethling-Graff, R.

2012-03-15

Highlights: Black-Right-Pointing-Pointer Two types of methanogens are necessary to respond successfully to perturbation. Black-Right-Pointing-Pointer Diversity of methanogens correlates with the VFA concentration and methane yield. Black-Right-Pointing-Pointer Aggregates indicate tight spatial relationship between minerals and microorganisms. - Abstract: Microbial community diversity in two thermophilic laboratory-scale and three full-scale anaerobic co-digesters was analysed by genetic profiling based on PCR-amplified partial 16S rRNA genes. In parallel operated laboratory reactors a stepwise increase of the organic loading rate (OLR) resulted in a decrease of methane production and an accumulation of volatile fatty acids (VFAs). However, almost threefold different OLRs were necessary to inhibit themore » gas production in the reactors. During stable reactor performance, no significant differences in the bacterial community structures were detected, except for in the archaeal communities. Sequencing of archaeal PCR products revealed a dominance of the acetoclastic methanogen Methanosarcina thermophila, while hydrogenotrophic methanogens were of minor importance and differed additionally in their abundance between reactors. As a consequence of the perturbation, changes in bacterial and archaeal populations were observed. After organic overload, hydrogenotrophic methanogens (Methanospirillum hungatei and Methanoculleus receptaculi) became more dominant, especially in the reactor attributed by a higher OLR capacity. In addition, aggregates composed of mineral and organic layers formed during organic overload and indicated tight spatial relationships between minerals and microbial processes that may support de-acidification processes in over-acidified sludge. Comparative analyses of mesophilic stationary phase full-scale reactors additionally indicated a correlation between the diversity of methanogens and the VFA concentration combined with the methane yield. This study demonstrates that the coexistence of two types of methanogens, i.e. hydrogenotrophic and acetoclastic methanogens is necessary to respond successfully to perturbation and leads to stable process performance.« less

Modelling of MOCVD Reactor: New 3D Approach

NASA Astrophysics Data System (ADS)

Raj, E.; Lisik, Z.; Niedzielski, P.; Ruta, L.; Turczynski, M.; Wang, X.; Waag, A.

2014-04-01

The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

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.

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.

DANSS Neutrino Spectrometer: Detector Calibration, Response Stability, and Light Yield

NASA Astrophysics Data System (ADS)

Alekseev, I. G.; Belov, V. V.; Danilov, M. V.; Zhitnikov, I. V.; Kobyakin, A. S.; Kuznetsov, A. S.; Machikhiliyan, I. V.; Medvedev, D. V.; Rusinov, V. Yu.; Svirida, D. N.; Skrobova, N. A.; Starostin, A. S.; Tarkovsky, E. I.; Fomina, M. V.; Shevchik, E. A.; Shirchenko, M. V.

2018-05-01

Apart from monitoring nuclear reactor parameters, the DANSS neutrino experiment is aimed at searching for sterile neutrinos through a detailed analysis of the ratio of reactor antineutrino spectra measured at different distances from the reactor core. The light collection system of the detector is dual, comprising both the vacuum photomultiplier tubes (PMTs) and silicon photomultipliers (SiPMs). In this paper, the techniques developed to calibrate the responses of these photodetectors are discussed in detail. The long-term stability of the key parameters of the detector and their dependences on the ambient temperature are investigated. The results of detector light yield measurements, performed independently with PMTs and SiPMs are reported.

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

None

Progress is reported on fundamental research in: crystal physics, reactions at metal surfaces, spectroscopy of ionic media, structure of metals, theory of alloying, physical properties, sintering, deformation of crystalline solids, x ray diffraction, metallurgy of superconducting materials, and electron microscope studies. Long-randge applied research studies were conducted for: zirconium metallurgy, materials compatibility, solid reactions, fuel element development, mechanical properties, non-destructive testing, and high-temperature materials. Reactor development support work was carried out for: gas-cooled reactor program, molten-salt reactor, high-flux isotope reactor, space-power program, thorium-utilization program, advanced-test reactor, Army Package Power Reactor, Enrico Fermi fast-breeder reactor, and water desalination program. Other programmore » activities, for which research was conducted, included: thermonuclear project, transuraniunn program, and post-irradiation examination laboratory. Separate abstracts were prepared for 30 sections of the report. (B.O.G.)« less

Control of Advanced Reactor-Coupled Heat Exchanger System: Incorporation of Reactor Dynamics in System Response to Load Disturbances

DOE PAGES

Skavdahl, Isaac; Utgikar, Vivek; Christensen, Richard; ...

2016-05-24

We present an alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) in this paper. One scheme is designed to control the cold outlet temperature of the SHX (T co) and the hot outlet temperature of the intermediate heat exchanger (T ho2) by manipulating the hot-side flow rates of the heat exchangers (F h/F h2) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the controlmore » of the cold outlet temperature of the SHX (T co) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The final option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.« less

NUCLEAR REACTORS

DOEpatents

Long, E.; Ashley, J.W.

1958-12-16

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

Shutdown-induced tensile stress in monolithic miniplates as a possible cause of plate pillowing at very high burnup

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

Medvedev, Pavel G; Ozaltun, Hakan; Robinson, Adam Brady

2014-04-01

Post-irradiation examination of Reduced Enrichment for Research and Test Reactors (RERTR)-12 miniplates showed that in-reactor pillowing occurred in at least 4 plates, rendering performance of these plates unacceptable. To address in-reactor failures, efforts are underway to define the mechanisms responsible for in-reactor pillowing, and to suggest improvements to the fuel plate design and operational conditions. To achieve these objectives, the mechanical response of monolithic fuel to fission and thermally-induced stresses was modeled using a commercial finite element analysis code. Calculations of stresses and deformations in monolithic miniplates during irradiation and after the shutdown revealed that the tensile stress generated inmore » the fuel increased from 2 MPa to 100 MPa at shutdown. The increase in tensile stress at shutdown possibly explains in-reactor pillowing of several RERTR-12 miniplates irradiated to the peak local burnup of up to 1.11x1022 fissions/cm3 . This paper presents the modeling approach and calculation results, and compares results with post-irradiation examinations and mechanical testing of irradiated fuel. The implications for the safe use of the monolithic fuel in research reactors are discussed, including the influence of fuel burnup and power on the magnitude of the shutdown-induced tensile stress.« less

Investigation of high temperature annealing effectiveness for recovery of radiation-induced structural changes and properties of 18Cr-10Ni-Ti austenitic stainless steels

NASA Astrophysics Data System (ADS)

Gurovich, B. A.; Kuleshova, E. A.; Frolov, A. S.; Maltsev, D. A.; Prikhodko, K. E.; Fedotova, S. V.; Margolin, B. Z.; Sorokin, A. A.

2015-10-01

A complex study of structural state and properties of 18Cr-10Ni-Ti austenitic stainless steel after irradiation in BOR-60 fast research reactor (in the temperature range 330-400 °С up to damaging doses of 145 dpa) and in VVER-1000 light water reactor (at temperature ∼320 °С and damaging doses ∼12-14 dpa) was performed. The possibility of recovery of structural-phase state and mechanical properties to the level almost corresponding to the initial state by the recovery annealing was studied. The principal possibility of the recovery annealing of pressurized water reactor internals that ensures almost complete recovery of its mechanical properties and microstructure was shown. The optimal mode of recovery annealing was established: 1000 °C during 120 h.

Science And Technology In Rapid Crisis Response. Volume 7, September 2011

DTIC Science & Technology

2011-09-01

Nuclear Reactor Accident at Fukushima Daiichi .... 16 From the Eyes of the Marines: III MEF’s Assistance in Japan Relief Efforts ............ 20 Chilean...by the radiological hazards at the Fukushima Daiichi Nuclear Power Plant. Although this project is in its early stages, the team hopes to have...OBSERVATIONS ON THE U.S. GOVERNMENT RESPONSE TO THE “GREAT EAST JAPAN EARTHQUAKE” AND THE NUCLEAR REACTOR ACCIDENT AT FUKUSHIMA DAIICHI CAPT Jim

Sloshing response of a reactor tank with internals

NASA Astrophysics Data System (ADS)

Ma, D. C.; Gvildys, J.; Chang, Y. W.

The sloshing response of a large reactor tank with in tank components is presented. It is indicated that the presence of the internal components can significantly change the dynamic characteristics of the sloshing motion. The sloshing frequency of a tank with internals is considerably higher than that of a tank without internal. The higher sloshing frequency reduces the sloshing wave height on the free surface but increases the dynamic pressure in the fluid.

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

Ivarson, Kristine A.; Miller, Charles W.; Arola, Craig C.

Groundwater contamination by hexavalent chromium and other nuclear reactor operation-related contaminants has resulted in the need for groundwater remedial actions within the Hanford Site reactor areas (the Hanford Site 100 Area). The large geographic extent of the resultant contaminant plumes requires an extensive level of understanding of the aquifer structure, characteristics, and configuration to support assessment and design of remedial alternatives within the former 100-D, 100-H, and 100-K reactor areas. The authors have prepared two- and three-dimensional depictions of the key subsurface geologic structures at two Hanford Site reactor operable units (100-K and 100-D/H). These depictions, prepared using commercial-off-the-shelf (COTS)more » visualization software, provide a basis for expanding the understanding of groundwater contaminant migration pathways, including identification of geologically-defined preferential groundwater flow pathways. These identified preferential flow pathways support the conceptual site model and help explain both historical and current contaminant distribution and transport. (authors)« less

Standardized reactors for the study of medical biofilms: a review of the principles and latest modifications.

PubMed

Gomes, Inês B; Meireles, Ana; Gonçalves, Ana L; Goeres, Darla M; Sjollema, Jelmer; Simões, Lúcia C; Simões, Manuel

2018-08-01

Biofilms can cause severe problems to human health due to the high tolerance to antimicrobials; consequently, biofilm science and technology constitutes an important research field. Growing a relevant biofilm in the laboratory provides insights into the basic understanding of the biofilm life cycle including responses to antibiotic therapies. Therefore, the selection of an appropriate biofilm reactor is a critical decision, necessary to obtain reproducible and reliable in vitro results. A reactor should be chosen based upon the study goals and a balance between the pros and cons associated with its use and operational conditions that are as similar as possible to the clinical setting. However, standardization in biofilm studies is rare. This review will focus on the four reactors (Calgary biofilm device, Center for Disease Control biofilm reactor, drip flow biofilm reactor, and rotating disk reactor) approved by a standard setting organization (ASTM International) for biofilm experiments and how researchers have modified these standardized reactors and associated protocols to improve the study and understanding of medical biofilms.

Cladding and duct materials for advanced nuclear recycle reactors

NASA Astrophysics Data System (ADS)

Allen, T. R.; Busby, J. T.; Klueh, R. L.; Maloy, S. A.; Toloczko, M. B.

2008-01-01

The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP’s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP’s advanced nuclear recycle reactors program.

Media arrangement impacts cell growth in anaerobic fixed-bed reactors treating sugarcane vinasse: Structured vs. randomic biomass immobilization.

PubMed

de Aquino, Samuel; Fuess, Lucas Tadeu; Pires, Eduardo Cleto

2017-07-01

This study reports on the application of an innovative structured-bed reactor (FVR) as an alternative to conventional packed-bed reactors (PBRs) to treat high-strength solid-rich wastewaters. Using the FVR prevents solids from accumulating within the fixed-bed, while maintaining the advantages of the biomass immobilization. The long-term operation (330days) of a FVR and a PBR applied to sugarcane vinasse under increasing organic loads (2.4-18.0kgCODm -3 day -1 ) was assessed, focusing on the impacts of the different media arrangements over the production and retention of biomass. Much higher organic matter degradation rates, as well as long-term operational stability and high conversion efficiencies (>80%) confirmed that the FVR performed better than the PBR. Despite the equivalent operating conditions, the biomass growth yield was different in both reactors, i.e., 0.095gVSSg -1 COD (FVR) and 0.066gVSSg -1 COD (PBR), indicating a clear control of the media arrangement over the biomass production in fixed-bed reactors. Copyright © 2017 Elsevier Ltd. All rights reserved.

Characterization of bacterial structures in two-stage moving-bed biofilm reactor (MBBR) during nitrification of the landfill leachate.

PubMed

Ciesielski, Slawomir; Kulikowska, Dorota; Kaczowka, Ewelina; Kowal, Przemysław

2010-07-01

Differences in DNA banding patterns, obtained by ribosomal intergenic spacer analysis (RISA), and nitrification were followed in a moving-bed biofilm reactor (MBBR) receiving municipal landfill leachate. Complete nitrification (> 99%) to nitrate was obtained in the two-stage MBBR system with an ammonium load of 1.09 g N-NH(4)/m(2).d. Increasing the ammonium load to 2.03 g N-NH(4)/m(2).d or more caused a decline in process efficiency to 70-86%. Moreover, at the highest ammonium load (3.76 g N-NH(4)/m(2).d), nitrite was the predominant product of nitrification. Community succession was evident in both compartments in response to changes in ammonium load. Non-metric multidimensional scaling (NMDS) supported by similarity analysis (ANOSIM) showed that microbial biofilm communities differed between compartments. The microbial biofilm was composed mainly of ammonia-oxidizing bacteria (AOB), with Nitrosomonas europeae and N. eutropha being most abundant. These results suggest that high ammonium concentrations select for particular AOB strains.

Comparative analysis of the bacterial diversity in a lab-scale moving bed biofilm reactor (MBBR) applied to treat urban wastewater under different operational conditions.

PubMed

Calderón, Kadiya; Martín-Pascual, Jaime; Poyatos, José Manuel; Rodelas, Belén; González-Martínez, Alejandro; González-López, Jesús

2012-10-01

Different types of carriers were tested as support material in a lab-scale moving bed biofilm reactor (MBBR) used to treat urban wastewater under three different conditions of hydraulic retention time (HRT) and carrier filling ratios (FR). The bacterial diversity developed on the biofilms responsible of the treatment was studied using a cultivation-independent approach based on the polymerase chain reaction-temperature gradient gel electrophoresis technique (PCR-TGGE). Cluster analysis of TGGE fingerprints showed significant differences of community structure dependent upon the different operational conditions applied. Redundancy analysis (RDA) was used to determine the relationship between the operational conditions (type of carrier, HRT, FR) and bacterial biofilm diversity, demonstrating a significant effect of FR=50%. Phylogenetic analysis of PCR-reamplified and sequenced TGGE bands revealed that the prevalent Bacteria populations in the biofilm were related to Betaproteobacteria (46%), Firmicutes (34%),Alphaproteobacteria (14%) and Gammaproteobacteria (9%). Copyright © 2012 Elsevier Ltd. All rights reserved.

Effect of mechanical disruption on the effectiveness of three reactors used for dilute acid pretreatment of corn stover Part 2: morphological and structural substrate analysis

PubMed Central

2014-01-01

Background Lignocellulosic biomass is a renewable, naturally mass-produced form of stored solar energy. Thermochemical pretreatment processes have been developed to address the challenge of biomass recalcitrance, however the optimization, cost reduction, and scalability of these processes remain as obstacles to the adoption of biofuel production processes at the industrial scale. In this study, we demonstrate that the type of reactor in which pretreatment is carried out can profoundly alter the micro- and nanostructure of the pretreated materials and dramatically affect the subsequent efficiency, and thus cost, of enzymatic conversion of cellulose. Results Multi-scale microscopy and quantitative image analysis was used to investigate the impact of different biomass pretreatment reactor configurations on plant cell wall structure. We identify correlations between enzymatic digestibility and geometric descriptors derived from the image data. Corn stover feedstock was pretreated under the same nominal conditions for dilute acid pretreatment (2.0 wt% H2SO4, 160°C, 5 min) using three representative types of reactors: ZipperClave® (ZC), steam gun (SG), and horizontal screw (HS) reactors. After 96 h of enzymatic digestion, biomass treated in the SG and HS reactors achieved much higher cellulose conversions, 88% and 95%, respectively, compared to the conversion obtained using the ZC reactor (68%). Imaging at the micro- and nanoscales revealed that the superior performance of the SG and HS reactors could be explained by reduced particle size, cellular dislocation, increased surface roughness, delamination, and nanofibrillation generated within the biomass particles during pretreatment. Conclusions Increased cellular dislocation, surface roughness, delamination, and nanofibrillation revealed by direct observation of the micro- and nanoscale change in accessibility explains the superior performance of reactors that augment pretreatment with physical energy. PMID:24690534

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

Response surface methodology modeling to improve degradation of Chlorpyrifos in agriculture runoff using TiO2 solar photocatalytic in a raceway pond reactor.

PubMed

Amiri, Hoda; Nabizadeh, Ramin; Silva Martinez, Susana; Jamaleddin Shahtaheri, Seyed; Yaghmaeian, Kamyar; Badiei, Alireza; Nazmara, Shahrokh; Naddafi, Kazem

2018-01-01

This paper deals with the use of a raceway pond reactor (RPR) as an alternative photoreactor for solar photocatalytic applications. Raceway pond reactors are common low-cost reactors which can treat large volumes of water. The experiments were carried out with TiO 2 in the agriculture effluent spiked with Chlorpyrifos (CPF) at circumneutral pH. The Response Surface Methodology (RSM) was used to find the optimum process parameters to maximize CPF oxidation from the mathematical model equations developed in this study using R software. By ANOVA, p-value of lack of fit > 0.05 indicated that, the equation was well-fitted. The theoretical efficiency of CPF removal, under the optimum oxidation conditions with UV solar energy of around 697 ± 5.33 lux, was 84.01%, which is in close agreement with the mean experimental value (80 ± 1.42%) confirming that the response model was suitable for the optimization. As far as the authors know, this is the first study of CPF removal using RPR in agriculture runoff at circumneutral pH. Copyright © 2017. Published by Elsevier Inc.

Testimony of Fred R. Mynatt before the Energy Research and Development Subcommittee of the Committee on Science, Space, and Technology, US House of Representatives. [Advanced fuel technology, gas-cooled reactor technology, and liquid metal-cooled reactor technology programs

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

Mynatt, F.R.

1987-03-18

This report provides a description of the statements submitted for the record to the committee on Science, Space, and Technology of the United States House of Representatives. These statements describe three principal areas of activity of the Advanced Reactor Technology Program of the Department of Energy (DOE). These areas are advanced fuel cycle technology, modular high-temperature gas-cooled reactor technology, and liquid metal-cooled reactor. The areas of automated reactor control systems, robotics, materials and structural design shielding and international cooperation were included in these statements describing the Oak Ridge National Laboratory's efforts in these areas. (FI)

A facility for testing 10 to 100-kWe space power reactors

NASA Astrophysics Data System (ADS)

Carlson, William F.; Bitten, Ernest J.

1993-01-01

This paper describes an existing facility that could be used in a cost-effective manner to test space power reactors in the 10 to 100-kWe range before launch. The facility has been designed to conduct full power tests of 100-kWe SP-100 reactor systems and already has the structural features that would be required for lower power testing. The paper describes a reasonable scenario starting with the acceptance at the test site of the unfueled reactor assembly and the separately shipped nuclear fuel. After fueling the reactor and installing it in the facility, cold critical tests are performed, and the reactor is then shipped to the launch site. The availability of this facility represents a cost-effective means of performing the required prelaunch test program.

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

Salko, Robert K; Sung, Yixing; Kucukboyaci, Vefa

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time stepmore » of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.« less

Neutronic Reactor Structure

DOEpatents

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

1961-05-30

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

NEUTRONIC REACTOR STRUCTURE

DOEpatents

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

1961-05-30

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

PRECAST CONCRETE WALL PANELS ARE LIFTED INTO PLACE ON MTR ...

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

PRECAST CONCRETE WALL PANELS ARE LIFTED INTO PLACE ON MTR STEEL FRAME STRUCTURE. INL NEGATIVE NO. 1330. Unknown Photographer, 1/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

Sklenka, L.; Rataj, J.; Frybort, J.

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

CONTROL FOR NEUTRONIC REACTOR

DOEpatents

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

1959-03-31

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

NEUTRONIC REACTOR CORE

DOEpatents

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

1961-07-18

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

A novel ultrasonic NDE for shrink fit welded structures using interface waves.

PubMed

Lee, Jaesun; Park, Junpil; Cho, Younho

2016-05-01

Reactor vessel inspection is a critical part of safety maintenance in a nuclear power plant. The inspection of shrink fit welded structures in a reactor nozzle can be a challenging task due to the complicated geometry. Nozzle inspection using pseudo interface waves allows us to inspect the nozzle from outside of the nuclear reactor. In this study, layered concentric pipes were manufactured with perfect shrink fit conditions using stainless steel 316. The displacement distributions were calculated with boundary conditions for a shrink fit welded structure. A multi-transducer guided wave phased array system was employed to monitor the welding quality of the nozzle end at a distance from a fixed position. The complicated geometry of a shrink fit welded structure can be overcome by using the pseudo interface waves in identifying the location and size of defects. The experimental results demonstrate the feasibility of detecting weld delamination and defects. Copyright © 2016 Elsevier B.V. All rights reserved.

Novel electrode structure in a DBD reactor applied to the degradation of phenol in aqueous solution

NASA Astrophysics Data System (ADS)

Mercado-Cabrera, Antonio; Peña-Eguiluz, Rosendo; López-Callejas, Régulo; Jaramillo-Sierra, Bethsabet; Valencia-Alvarado, Raúl; Rodríguez-Méndez, Benjamín; Muñoz-Castro, Arturo E.

2017-07-01

Phenol degradation experimental results are presented in a similar wastewater aqueous solution using a non-thermal plasma reactor in a coaxial dielectric barrier discharge. The novelty of the work is that one of the electrodes of the reactor has the shape of a hollow screw which shows an enhanced efficiency compared with a traditional smooth structure. The experimentation was carried out with gas mixtures of 90% Ar-10% O2, 80% Ar-20% O2 and 0% Ar-100% O2. After one hour of treatment the removal efficiency was 76%, 92%, and 97%, respectively, assessed with a gas chromatographic mass spectrometry technique. For both reactors used, the ozone concentration was measured. The screw electrode required less energy, for all gas mixtures, than the smooth electrode, to maintain the same ozone concentration. On the other hand, it was also observed that in both electrodes the electrical conductivity of the solution changed slightly from ˜0.0115 S m-1 up to ˜0.0430 S m-1 after one hour of treatment. The advantages of using the hollow screw electrode structure compared with the smooth electrode were: (1) lower typical power consumption, (2) the generation of a uniform plasma throughout the reactor benefiting the phenol degradation, (3) a relatively lower temperature of the aqueous solution during the process, and (4) the plasma generation length is larger.

Response And Recovery Of Sulfate-Reducing Biochemical Reactors From Aerobic Stress Events

EPA Science Inventory

Microbially-mediated treatment of mining-influenced water (MIW) through the implementation of sulfate-reducing biochemical reactors (BCRs) is an attractive option for passive, in situ remediation with low operating costs and reduced maintenance requirements. However, BCRs can be...

Response And Recovery Of Sulfate-Reducing Biochemical Reactors From Aerobic Stress Events (Presentation)

EPA Science Inventory

Microbially-mediated treatment of mining-influenced water (MIW) through the implementation of sulfate-reducing biochemical reactors (BCR) is an attractive option for passive, in situ remediation with low operating costs and reduced maintenance requirements. However, BCRs can be ...

Catalytic reactor for promoting a chemical reaction on a fluid passing therethrough

NASA Technical Reports Server (NTRS)

Roychoudhury, Subir (Inventor); Pfefferle, William C. (Inventor)

2001-01-01

A catalytic reactor with an auxiliary heating structure for raising the temperature of a fluid passing therethrough whereby the catalytic reaction is promoted. The invention is a apparatus employing multiple electrical heating elements electrically isolated from one another by insulators that are an integral part of the flow path. The invention provides step heating of a fluid as the fluid passes through the reactor.

Reactor for tracking catalyst nanoparticles in liquid at high temperature under a high-pressure gas phase with X-ray absorption spectroscopy.

PubMed

Nguyen, Luan; Tao, Franklin Feng

2018-02-01

Structure of catalyst nanoparticles dispersed in liquid phase at high temperature under gas phase of reactant(s) at higher pressure (≥5 bars) is important for fundamental understanding of catalytic reactions performed on these catalyst nanoparticles. Most structural characterizations of a catalyst performing catalysis in liquid at high temperature under gas phase at high pressure were performed in an ex situ condition in terms of characterizations before or after catalysis since, from technical point of view, access to the catalyst nanoparticles during catalysis in liquid phase at high temperature under high pressure reactant gas is challenging. Here we designed a reactor which allows us to perform structural characterization using X-ray absorption spectroscopy including X-ray absorption near edge structure spectroscopy and extended X-ray absorption fine structure spectroscopy to study catalyst nanoparticles under harsh catalysis conditions in terms of liquid up to 350 °C under gas phase with a pressure up to 50 bars. This reactor remains nanoparticles of a catalyst homogeneously dispersed in liquid during catalysis and X-ray absorption spectroscopy characterization.

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.

Impact of Substratum Surface on Microbial Community Structure and Treatment Performance in Biological Aerated Filters

PubMed Central

Kim, Lavane; Pagaling, Eulyn; Zuo, Yi Y.

2014-01-01

The impact of substratum surface property change on biofilm community structure was investigated using laboratory biological aerated filter (BAF) reactors and molecular microbial community analysis. Two substratum surfaces that differed in surface properties were created via surface coating and used to develop biofilms in test (modified surface) and control (original surface) BAF reactors. Microbial community analysis by 16S rRNA gene-based PCR-denaturing gradient gel electrophoresis (DGGE) showed that the surface property change consistently resulted in distinct profiles of microbial populations during replicate reactor start-ups. Pyrosequencing of the bar-coded 16S rRNA gene amplicons surveyed more than 90% of the microbial diversity in the microbial communities and identified 72 unique bacterial species within 19 bacterial orders. Among the 19 orders of bacteria detected, Burkholderiales and Rhodocyclales of the Betaproteobacteria class were numerically dominant and accounted for 90.5 to 97.4% of the sequence reads, and their relative abundances in the test and control BAF reactors were different in consistent patterns during the two reactor start-ups. Three of the five dominant bacterial species also showed consistent relative abundance changes between the test and control BAF reactors. The different biofilm microbial communities led to different treatment efficiencies, with consistently higher total organic carbon (TOC) removal in the test reactor than in the control reactor. Further understanding of how surface properties affect biofilm microbial communities and functional performance would enable the rational design of new generations of substrata for the improvement of biofilm-based biological treatment processes. PMID:24141134

Operators in the Plum Brook Reactor Facility Control Room

NASA Image and Video Library

1970-03-21

Donald Rhodes, left, and Clyde Greer, right, monitor the operation of the National Aeronautics and Space Administration’s (NASA) Plum Brook Reactor Facility from the control room. The 60-megawatt test reactor, NASA’s only reactor, was the eighth largest test reactor in the world. The facility was built by the Lewis Research Center in the late 1950s to study the effects of radiation on different materials that could be used to construct nuclear propulsion systems for aircraft or rockets. The reactor went critical for the first time in 1961. For the next two years, two operators were on duty 24 hours per day working on the fission process until the reactor reached its full-power level in 1963. Reactor Operators were responsible for monitoring and controlling the reactor systems. Once the reactor was running under normal operating conditions, the work was relatively uneventful. Normally the reactor was kept at a designated power level within certain limits. Occasionally the operators had to increase the power for a certain test. The shift supervisor and several different people would get together and discuss the change before boosting the power. All operators were required to maintain a Reactor Operator License from the Atomic Energy Commission. The license included six months of training, an eight-hour written exam, a four-hour walkaround, and testing on the reactor controls.

Advanced sample environments for in situ neutron diffraction studies of nuclear materials

NASA Astrophysics Data System (ADS)

Reiche, Helmut Matthias

Generation IV nuclear reactor concepts, such as the supercritical-water-cooled nuclear reactor (SCWR), are actively researched internationally. Operating conditions above the critical point of water (374°C, 22.1 MPa) and fuel core temperature that potentially exceed 1850°C put a high demand on the surrounding materials. For their safe application, it is essential to characterize and understand the material properties on an atomic scale such as crystal structure and grain orientation (texture) changes as a function of temperature and stress. This permits the refinement of models predicting the macroscopic behavior of the material. Neutron diffraction is a powerful tool in characterizing such crystallographic properties due to their deep penetration depth into condensed matter. This leads to the ability to study bulk material properties, as opposed to surface effects, and allows for complex sample environments to study e.g. the individual contributions of thermo-mechanical processing steps during manufacturing, operating or accident scenarios. I present three sample environments for in situ neutron diffraction studies that provide such crystallographic information and have been successfully commissioned and integrated into the user program of the High Pressure -- Preferred Orientation (HIPPO) diffractometer at the Los Alamos Neutron Science Center (LANSCE) user facility. I adapted a sample changer for reliable and fast automated texture measurements of multiple specimens. I built a creep furnace combining a 2700 N load frame with a resistive vanadium furnace, capable of temperatures up to 1000°C, and manipulated by a pair of synchronized rotation stages. This combination allows following deformation and temperature dependent texture and strain evolutions in situ. Utilizing the presented sample changer and creep furnace we studied pressure tubes made of Zr-2.5wt%Nb currently employed in CANDURTM nuclear reactors and proposed for future SCWRs, acting as the primary containment vessel of high temperature heavy water (D2O) inside the reactor core. The measured sample texture shows that upon traversing the phase transition, which proceeded according to the Burger orientation relationship, variant selection occurred during heating and cooling of the zirconium alloy. Experimental results of lattice strains depending on the crystallographic orientation can be used to calculate strain pole figures which grant insight into the three-dimensional mechanical response of a polycrystalline aggregate and represent an extremely powerful material model validation tool. Lastly, I developed a resistive graphite high-temperature furnace with sample motion for in situ crystal structure and texture measurements of nuclear materials at steady-state temperatures up to at least 2200°C. This permits in situ observation of e.g. phase transitions and coefficients of thermal expansion, as well as phase formation and texture development during solidification. Utilizing this apparatus, I investigated the carbothermic reduction of UO2 nanopowder forming uranium carbide, a promising Generation IV reactor fuel. The onset of the UO2 + 2C → UC + CO2 reaction was observed at 1440°C with the bulk portion being complete at 1500°C. I describe the novel synthesis for this nanoparticle UO2 powder, which closely imitates observed nano grains in partially burnt reactor fuels. Of the three opposing structure models reported for the non-quenchable cubic UC2 phase, stable between 1769°C and 2560°C, the NaCl-type structure according to Bowman is found to be correct. This is deemed major progress as the CaF2-type structure was used for recent thermal modeling of safety critical factors in nuclear reactors. A temperature dependent increase in density due to carbon diffusion has been observed and quantified. I provide first experimental data of an unspecified, reversible order-disorder transition in this delta-phase with its onset at ˜1800°C which is likely due to rotating C2 molecules in the sublattice.

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

Effect of Different Structural Materials on Neutronic Performance of a Hybrid Reactor

NASA Astrophysics Data System (ADS)

Übeyli, Mustafa; Tel, Eyyüp

2003-06-01

Selection of structural material for a fusion-fission (hybrid) reactor is very important by taking into account of neutronic performance of the blanket. Refractory metals and alloys have much higher operating temperatures and neutron wall load (NWL) capabilities than low activation materials (ferritic/martensitic steels, vanadium alloys and SiC/SiC composites) and austenitic stainless steels. In this study, effect of primary candidate refractory alloys, namely, W-5Re, T111, TZM and Nb-1Zr on neutronic performance of the hybrid reactor was investigated. Neutron transport calculations were conducted with the help of SCALE 4.3 System by solving the Boltzmann transport equation with code XSDRNPM. Among the investigated structural materials, tantalum had the worst performance due to the fact that it has higher neutron absorption cross section than others. And W-5Re and TZM having similar results showed the best performance.

Advanced low-activation materials. Fibre-reinforced ceramic composites

NASA Astrophysics Data System (ADS)

Fenici, P.; Scholz, H. W.

1994-09-01

A serious safety and environmental concern for thermonuclear fusion reactor development regards the induced radioactivity of the first wall and structural components. The use of low-activation materials (LAM) in a demonstration reactor would reduce considerably its potential risk and facilitate its maintenance. Moreover, decommissioning and waste management including disposal or even recycling of structural materials would be simplified. Ceramic fibre-reinforced SiC materials offer highly appreciable low activation characteristics in combination with good thermomechanical properties. This class of materials is now under experimental investigation for structural application in future fusion reactors. An overview on the recent results is given, covering coolant leak rates, thermophysical properties, compatibility with tritium breeder materials, irradiation effects, and LAM-consistent purity. SiC/SiC materials present characteristics likely to be optimised in order to meet the fusion application challenge. The scope is to put into practice the enormous potential of inherent safety with fusion energy.

Neutron-gamma flux and dose calculations for feasibility study of DISCOMS instrumentation in case of severe accident in a GEN 3 reactor

NASA Astrophysics Data System (ADS)

Brovchenko, Mariya; Duhamel, Isabelle; Dechenaux, Benjamin

2017-09-01

The present paper presents the study carried out in the frame of the DISCOMS project, which stands for "DIstributed Sensing for COrium Monitoring and Safety". This study concerns the calculation of the neutron and gamma radiations received by the considered instrumentation during the normal reactor operation as well as in case of a severe accident for the EPR reactor, outside the reactor pressure vessel and in the containment basemat. This paper summarizes the methods and hypotheses used for the particle transport simulation outside the vessel during normal reactor operation. The results of the simulations are then presented including the responses for distributed Optical Fiber Sensors (OFS), such as the gamma dose and the fast neutron fluence, and for Self Powered Neutron Detectors (SPNDs), namely the neutron and gamma spectra. Same responses are also evaluated for severe accident situations in order to design the SPNDs being sensitive to the both types of received neutron-gamma radiation. By contrast, fibers, involved as transducers in distributed OFS have to resist to the total radiation gamma dose and neutron fluence received during normal operation and the severe accident.

Process optimization of an auger pyrolyzer with heat carrier using response surface methodology.

PubMed

Brown, J N; Brown, R C

2012-01-01

A 1 kg/h auger reactor utilizing mechanical mixing of steel shot heat carrier was used to pyrolyze red oak wood biomass. Response surface methodology was employed using a circumscribed central composite design of experiments to optimize the system. Factors investigated were: heat carrier inlet temperature and mass flow rate, rotational speed of screws in the reactor, and volumetric flow rate of sweep gas. Conditions for maximum bio-oil and minimum char yields were high flow rate of sweep gas (3.5 standard L/min), high heat carrier temperature (∼600 °C), high auger speeds (63 RPM) and high heat carrier mass flow rates (18 kg/h). Regression models for bio-oil and char yields are described including identification of a novel interaction effect between heat carrier mass flow rate and auger speed. Results suggest that auger reactors, which are rarely described in literature, are well suited for bio-oil production. The reactor achieved liquid yields greater than 73 wt.%. Copyright © 2011 Elsevier Ltd. All rights reserved.

Microbial community functional structure in response to micro-aerobic conditions in sulfate-reducing sulfur-producing bioreactor.

PubMed

Yu, Hao; Chen, Chuan; Ma, Jincai; Xu, Xijun; Fan, Ronggui; Wang, Aijie

2014-05-01

Limited oxygen supply to anaerobic wastewater treatment systems had been demonstrated as an effective strategy to improve elemental sulfur (S(0)) recovery, coupling sulfate reduction and sulfide oxidation. However, little is known about the impact of dissolved oxygen (DO) on the microbial functional structures in these systems. We used a high throughput tool (GeoChip) to evaluate the microbial community structures in a biological desulfurization reactor under micro-aerobic conditions (DO: 0.02-0.33 mg/L). The results indicated that the microbial community functional compositions and structures were dramatically altered with elevated DO levels. The abundances of dsrA/B genes involved in sulfate reduction processes significantly decreased (p < 0.05, LSD test) at relatively high DO concentration (DO: 0.33 mg/L). The abundances of sox and fccA/B genes involved in sulfur/sulfide oxidation processes significantly increased (p < 0.05, LSD test) in low DO concentration conditions (DO: 0.09 mg/L) and then gradually decreased with continuously elevated DO levels. Their abundances coincided with the change of sulfate removal efficiencies and elemental sulfur (S(0)) conversion efficiencies in the bioreactor. In addition, the abundance of carbon degradation genes increased with the raising of DO levels, showing that the heterotrophic microorganisms (e.g., fermentative microorganisms) were thriving under micro-aerobic condition. This study provides new insights into the impacts of micro-aerobic conditions on the microbial functional structure of sulfate-reducing sulfur-producing bioreactors, and revealed the potential linkage between functional microbial communities and reactor performance. Copyright © 2014 The Research Centre for Eco-Environmental Sciences, Chinese Academy of Sciences. Published by Elsevier B.V. All rights reserved.

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

Molecular characterization of anaerobic sulfur-oxidizing microbial communities in up-flow anaerobic sludge blanket reactor treating municipal sewage.

PubMed

Aida, Azrina A; Hatamoto, Masashi; Yamamoto, Masamitsu; Ono, Shinya; Nakamura, Akinobu; Takahashi, Masanobu; Yamaguchi, Takashi

2014-11-01

A novel wastewater treatment system consisting of an up-flow anaerobic sludge blanket (UASB) reactor and a down-flow hanging sponge (DHS) reactor with sulfur-redox reaction was developed for treatment of municipal sewage under low-temperature conditions. In the UASB reactor, a novel phenomenon of anaerobic sulfur oxidation occurred in the absence of oxygen, nitrite and nitrate as electron acceptors. The microorganisms involved in anaerobic sulfur oxidation have not been elucidated. Therefore, in this study, we studied the microbial communities existing in the UASB reactor that probably enhanced anaerobic sulfur oxidation. Sludge samples collected from the UASB reactor before and after sulfur oxidation were used for cloning and terminal restriction fragment length polymorphism (T-RFLP) analysis of the 16S rRNA genes of the bacterial and archaeal domains. The microbial community structures of bacteria and archaea indicated that the genus Smithella and uncultured bacteria within the phylum Caldiserica were the dominant bacteria groups. Methanosaeta spp. was the dominant group of the domain archaea. The T-RFLP analysis, which was consistent with the cloning results, also yielded characteristic fingerprints for bacterial communities, whereas the archaeal community structure yielded stable microbial community. From these results, it can be presumed that these major bacteria groups, genus Smithella and uncultured bacteria within the phylum Caldiserica, probably play an important role in sulfur oxidation in UASB reactors. Copyright © 2014 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

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.

Engineering and Fabrication Considerations for Cost-Effective Space Reactor Shield Development

NASA Astrophysics Data System (ADS)

Berg, Thomas A.; Disney, Richard K.

2004-02-01

Investment in developing nuclear power for space missions cannot be made on the basis of a single mission. Current efforts in the design and fabrication of the reactor module, including the reactor shield, must be cost-effective and take into account scalability and fabricability for planned and future missions. Engineering considerations for the shield need to accommodate passive thermal management, varying radiation levels and effects, and structural/mechanical issues. Considering these challenges, design principles and cost drivers specific to the engineering and fabrication of the reactor shield are presented that contribute to lower recurring mission costs.

Engineering and Fabrication Considerations for Cost-Effective Space Reactor Shield Development

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

Berg, Thomas A.; Disney, Richard K.

Investment in developing nuclear power for space missions cannot be made on the basis of a single mission. Current efforts in the design and fabrication of the reactor module, including the reactor shield, must be cost-effective and take into account scalability and fabricability for planned and future missions. Engineering considerations for the shield need to accommodate passive thermal management, varying radiation levels and effects, and structural/mechanical issues. Considering these challenges, design principles and cost drivers specific to the engineering and fabrication of the reactor shield are presented that contribute to lower recurring mission costs.

Emergency heat removal system for a nuclear reactor

DOEpatents

Dunckel, Thomas L.

1976-01-01

A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

JOYO-1 Irradiation Test Campaign Technical Close-out, For Information

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

G. Borges

2006-01-31

The JOYO-1 irradiation testing was designed to screen the irradiation performance of candidate cladding, structural and reflector materials in support of space reactor development. The JOYO-1 designation refers to the first of four planned irradiation tests in the JOYO reactor. Limited irradiated material performance data for the candidate materials exists for the expected Prometheus-1 duration, fluences and temperatures. Materials of interest include fuel element cladding and core materials (refractory metal alloys and silicon carbide (Sic)), vessel and plant structural materials (refractory metal alloys and nickel-base superalloys), and control and reflector materials (BeO). Key issues to be evaluated were long termmore » microstructure and material property stability. The JOYO-1 test campaign was initiated to irradiate a matrix of specimens at prototypical temperatures and fluences anticipated for the Prometheus-1 reactor [Reference (1)]. Enclosures 1 through 9 describe the specimen and temperature monitors/dosimetry fabrication efforts, capsule design, disposition of structural material irradiation rigs, and plans for post-irradiation examination. These enclosures provide a detailed overview of Naval Reactors Prime Contractor Team (NRPCT) progress in specific areas; however, efforts were in various states of completion at the termination of NRPCT involvement with and restructuring of Project Prometheus.« less

Nuclear design of a very-low-activation fusion reactor

NASA Astrophysics Data System (ADS)

Cheng, E. T.; Hopkins, G. R.

1983-06-01

The nuclear design aspects of using very-low-activation materials, such as SiC, MgO, and aluminum for fusion-reactor first wall, blanket, and shield applications were investigated. In addition to the advantage of very-low radioactive inventory, it was found that the very-low-activation fusion reactor can also offer an adequate tritium-breeding ratio and substantial amount of blanket nuclear heating as a conventional-material-structured reactor does. The most-stringent design constraint found in a very-low-activation fusion reactor is the limited space available in the inboard region of a Tokamak concept for shielding to protect the superconducting toroidal field coil. A reference design was developed which mitigates the constraint by adopting a removable tungsten shield design that retains the inboard dimensions and gives the same shield performance as the reference STARFIRE Tokamak reactor design.

Membrane biofouling mechanism in an aerobic granular reactor degrading 4-chlorophenol.

PubMed

Buitrón, Germán; Moreno-Andrade, Iván; Arellano-Badillo, Víctor M; Ramírez-Amaya, Víctor

2014-01-01

The membrane fouling of an aerobic granular reactor coupled with a submerged membrane in a sequencing batch reactor (SBR) was evaluated. The fouling analysis was performed by applying microscopy techniques to determine the morphology and structure of the fouling layer on a polyvinylidene fluoride membrane. It was found that the main cause of fouling was the polysaccharide adsorption on the membrane surface, followed by the growth of microorganisms to form a biofilm.

The 5-kwe reactor thermoelectric system summary

NASA Technical Reports Server (NTRS)

Vanosdol, J. H. (Editor)

1973-01-01

Design of the 5-kwe reactor thermoelectric system was initiated in February 1972 and extended through the conceptual design phase into the preliminary design phase. Design effort was terminated in January, 1973. This report documents the system and component requirements, design approaches, and performance and design characteristics for the 5-kwe system. Included is summary information on the reactor, radiation shields, power conversion systems, thermoelectric pump, radiator/structure, liquid metal components, and the control system.

SURROGATE MODEL DEVELOPMENT AND VALIDATION FOR RELIABILITY ANALYSIS OF REACTOR PRESSURE VESSELS

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

Hoffman, William M.; Riley, Matthew E.; Spencer, Benjamin W.

In nuclear light water reactors (LWRs), the reactor coolant, core and shroud are contained within a massive, thick walled steel vessel known as a reactor pressure vessel (RPV). Given the tremendous size of these structures, RPVs typically contain a large population of pre-existing flaws introduced in the manufacturing process. After many years of operation, irradiation-induced embrittlement makes these vessels increasingly susceptible to fracture initiation at the locations of the pre-existing flaws. Because of the uncertainty in the loading conditions, flaw characteristics and material properties, probabilistic methods are widely accepted and used in assessing RPV integrity. The Fracture Analysis of Vesselsmore » – Oak Ridge (FAVOR) computer program developed by researchers at Oak Ridge National Laboratory is widely used for this purpose. This program can be used in order to perform deterministic and probabilistic risk-informed analyses of the structural integrity of an RPV subjected to a range of thermal-hydraulic events. FAVOR uses a one-dimensional representation of the global response of the RPV, which is appropriate for the beltline region, which experiences the most embrittlement, and employs an influence coefficient technique to rapidly compute stress intensity factors for axis-aligned surface-breaking flaws. The Grizzly code is currently under development at Idaho National Laboratory (INL) to be used as a general multiphysics simulation tool to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled RPVs. Grizzly can be used to model the thermo-mechanical response of an RPV under transient conditions observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local 3D models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which can in turn be used to assess whether a fracture would initiate at a pre-existing flaw. To use Grizzly for probabilistic analysis, it is necessary to have a way to rapidly evaluate stress intensity factors. To accomplish this goal, a reduced order model (ROM) has been developed to efficiently represent the behavior of a detailed 3D Grizzly model used to calculate fracture parameters. This approach uses the stress intensity factor influence coefficient method that has been used with great success in FAVOR. Instead of interpolating between tabulated solutions, as FAVOR does, the ROM approach uses a response surface methodology to compute fracture solutions based on a sampled set of results used to train the ROM. The main advantages of this approach are that the process of generating the training data can be fully automated, and the procedure can be readily used to consider more general flaw configurations. This paper demonstrates the procedure used to generate a ROM to rapidly compute stress intensity factors for axis-aligned flaws. The results from this procedure are in good agreement with those produced using the traditional influence coefficient interpolation procedure, which gives confidence in this method. This paves the way for applying this procedure for more general flaw configurations.« less

10 CFR 50.48 - Fire protection.

Code of Federal Regulations, 2011 CFR

2011-01-01

... suppression systems; and (iii) The means to limit fire damage to structures, systems, or components important...) Standard 805, “Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating... pressurized-water reactors (PWRs) is not permitted. (iv) Uncertainty analysis. An uncertainty analysis...

10 CFR 50.48 - Fire protection.

Code of Federal Regulations, 2010 CFR

2010-01-01

... suppression systems; and (iii) The means to limit fire damage to structures, systems, or components important...) Standard 805, “Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating... pressurized-water reactors (PWRs) is not permitted. (iv) Uncertainty analysis. An uncertainty analysis...

10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2014 CFR

2014-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2012 CFR

2012-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2011 CFR

2011-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

10 CFR 1.32 - Office of the Executive Director for Operations.

Code of Federal Regulations, 2013 CFR

2013-01-01

... of Nuclear Reactor Regulation, the Office of New Reactors, the Office of Nuclear Material Safety and... Section 1.32 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION... Nuclear Regulatory Research, the Office of Nuclear Security and Incident Response, and the NRC Regional...

Advanced reactor design study. Assessing nonbackfittable concepts for improving uranium utilization in light water reactors

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

Fleischman, R.M.; Goldsmith, S.; Newman, D.F.

1981-09-01

The objective of the Advanced Reactor Design Study (ARDS) is to identify and evaluate nonbackfittable concepts for improving uranium utilization in light water reactors (LWRs). The results of this study provide a basis for selecting and demonstrating specific nonbackfittable concepts that have good potential for implementation. Lead responsibility for managing the study was assigned to the Pacific Northwest Laboratory (PNL). Nonbackfittable concepts for improving uranium utilization in LWRs on the once-through fuel cycle were selected separately for PWRs and BWRs due to basic differences in the way specific concepts apply to those plants. Nonbackfittable concepts are those that are toomore » costly to incorporate in existing plants, and thus, could only be economically incorporated in new reactor designs or plants in very early stages of construction. Essential results of the Advanced Reactor Design Study are summarized.« less

Neutron Radiation Damage Estimation in the Core Structure Base Metal of RSG GAS

NASA Astrophysics Data System (ADS)

Santa, S. A.; Suwoto

2018-02-01

Radiation damage in core structure of the Indonesian RGS GAS multi purpose reactor resulting from the reaction of fast and thermal neutrons with core material structure was investigated for the first time after almost 30 years in operation. The aim is to analyze the degradation level of the critical components of the RSG GAS reactor so that the remaining life of its component can be estimated. Evaluation results of critical components remaining life will be used as data ccompleteness for submission of reactor operating permit extension. Material damage analysis due to neutron radiation is performed for the core structure components made of AlMg3 material and bolts reinforcement of core structure made of SUS304. Material damage evaluation was done on Al and Fe as base metal of AlMg3 and SUS304, respectively. Neutron fluences are evaluated based on the assumption that neutron flux calculations of U3Si8-Al equilibrium core which is operated on power rated of 15 MW. Calculation result using SRAC2006 code of CITATION module shows the maximum total neutron flux and flux >0.1 MeV are 2.537E+14 n/cm2/s and 3.376E+13 n/cm2/s, respectively. It was located at CIP core center close to the fuel element. After operating up to the end of #89 core formation, the total neutron fluence and fluence >0.1 MeV were achieved 9.063E+22 and 1.269E+22 n/cm2, respectively. Those are related to material damage of Al and Fe as much as 17.91 and 10.06 dpa, respectively. Referring to the life time of Al-1100 material irradiated in the neutron field with thermal flux/total flux=1.7 which capable of accepting material damage up to 250 dpa, it was concluded that RSG GAS reactor core structure underwent 7.16% of its operating life span. It means that core structure of RSG GAS reactor is still capable to receive the total neutron fluence of 9.637E+22 n/cm2 or fluence >0.1 MeV of 5.672E+22 n/cm2.

REACTOR UNLOADING

DOEpatents

Leverett, M.C.

1958-02-18

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

Development of the Technology of Vortex Diagnostics to Improve the Safety of Operation of Nuclear Reactors

NASA Astrophysics Data System (ADS)

Mitrofanova, O. V.; Ivlev, O. A.; Pozdeeva, I. G.; Urtenov, D. S.

2017-11-01

The results of studies are aimed at developing theoretical foundations and instrumentation system to ensure a technology of vortex diagnostics of the state of flows of fluids for nuclear power installations with power water reactors and fast neutrons reactors with liquid-metal coolants. The technology of vortex diagnostics is based on the study of acoustic, magneto-hydrodynamic and resonant effects related to the formation of stable vortex structures. For creation a system of monitoring and diagnostics of the crisis phenomena due to hydrodynamics of the flow, it is proposed to use acoustic method to record the radiation of elastic waves in the fluids caused by the dynamic local rearrangement of its structure.

ENGINEERING TEST REACTOR, TRA642. CONTEXTUAL VIEW ORIENTATING ETR TO MTR. ...

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

ENGINEERING TEST REACTOR, TRA-642. CONTEXTUAL VIEW ORIENTATING ETR TO MTR. CAMERA IS ON ROOF OF MTR BUILDING AND FACES DUE SOUTH. MTR SERVICE BUILDING, TRA-635, IN LOWER RIGHT CORNER. STEEL FRAMES SHOW BUILDINGS TO BE ATTACHED TO ETR BUILDING. HIGH-BAY SECTION IN CENTER IS REACTOR BUILDING. TWO-STORY CONTROL ROOM AND OFFICE BUILDING, TRA-647, IS BETWEEN IT AND MTR SERVICE BUILDING. STRUCTURE TO THE LEFT (WITH NO FRAMING YET) IS COMPRESSOR BUILDING, TRA-643, AND BEYOND IT WILL BE HEAT EXCHANGER BUILDING, TRA-644, GREAT SOUTHERN BUTTE ON HORIZON. INL NEGATIVE NO. 56-2382. Jack L. Anderson, Photographer, 6/10/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Physical and chemical controls on the critical zone

USGS Publications Warehouse

Anderson, S.P.; Von Blanckenburg, F.; White, A.F.

2007-01-01

Geochemists have long recognized a correlation between rates of physical denudation and chemical weathering. What underlies this correlation? The Critical Zone can be considered as a feed-through reactor. Downward advance of the weathering front brings unweathered rock into the reactor. Fluids are supplied through precipitation. The reactor is stirred at the top by biological and physical processes. The balance between advance of the weathering front by mechanical and chemical processes and mass loss by denudation fixes the thickness of the Critical Zone reactor. The internal structure of this reactor is controlled by physical processes that create surface area, determine flow paths, and set the residence time of material in the Critical Zone. All of these impact chemical weathering flux.

Down-selection of candidate alloys for further testing of advanced replacement materials for LWR core internals

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

Was, Gary; Leonard, Keith J.; Tan, Lizhen

Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) Light Water Reactor Sustainability Program to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to identify and develop advanced alloys with superiormore » degradation resistance in light water reactor (LWR)-relevant environments by 2024.« less

Characterization of gamma field in the JSI TRIGA reactor

NASA Astrophysics Data System (ADS)

Ambrožič, Klemen; Radulović, Vladimir; Snoj, Luka; Gruel, Adrien; Guillou, Mael Le; Blaise, Patrick; Destouches, Christophe; Barbot, Loïc

2018-01-01

Research reactors such as the "Jožzef Stefan" Institute TRIGA reactor have primarily been designed for experimentation and sample irradiation with neutrons. However recent developments in incorporating additional instrumentation for nuclear power plant support and with novel high flux material testing reactor designs, γ field characterization has become of great interest for the characterization of the changes in operational parameters of electronic devices and for the evaluation of γ heating of MTR's structural materials in a representative reactor Γ spectrum. In this paper, we present ongoing work on γ field characterization both experimentally, by performing γ field measurements, and by simulations, using Monte Carlo particle transport codes in conjunction with R2S methodology for delayed γ field characterization.

Application of Reactor Antineutrinos: Neutrinos for Peace

NASA Astrophysics Data System (ADS)

Suekane, F.

2013-02-01

In nuclear reactors, 239Pu are produced along with burn-up of nuclear fuel. 239Pu is subject of safeguard controls since it is an explosive component of nuclear weapon. International Atomic Energy Agency (IAEA) is watching undeclared operation of reactors to prevent illegal production and removal of 239Pu. In operating reactors, a huge numbers of anti electron neutrinos (ν) are produced. Neutrino flux is approximately proportional to the operating power of reactor in short term and long term decrease of the neutrino flux per thermal power is proportional to the amount of 239Pu produced. Thus rector ν's carry direct and real time information useful for the safeguard purposes. Since ν can not be hidden, it could be an ideal medium to monitor the reactor operation. IAEA seeks for novel technologies which enhance their ability and reactor neutrino monitoring is listed as one of such candidates. Currently neutrino physicists are performing R&D of small reactor neutrino detectors to use specifically for the safeguard use in response to the IAEA interest. In this proceedings of the neutrino2012 conference, possibilities of such reactor neutrinos application and current world-wide R&D status are described.

Nuclear reactor shutdown control rod assembly

DOEpatents

Bilibin, Konstantin

1988-01-01

A temperature responsive, self-actuated nuclear reactor shutdown control rod assembly 10. The upper end 18 of a lower drive line 17 fits within the lower end of an upper drive line 12. The lower end (not shown) of the lower drive line 17 is connected to a neutron absorber. During normal temperature conditions the lower drive line 17 is supported by detent means 22,26. When an overtemperature condition occurs thermal actuation means 34 urges ring 26 upwardly sufficiently to allow balls 22 to move radially outwardly thereby allowing lower drive line 17 to move downwardly toward the core of the nuclear reactor resulting in automatic reduction of the reactor powder.

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

Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.

As part of the Phase II testing at the HDR Test Facility in Kahl/Main, FRG, two series of high-level seismic/vibrational experiments were performed. In the first of these (SHAG) a coast-down shaker, mounted on the reactor operating floor and capable of generating 1000 tonnes of force, was used to investigate full-scale structural response, soil-structure interaction (SSI), and piping/equipment response at load levels equivalent to those of a design basis earthquake. The HDR soil/structure system was tested to incipient failure exhibiting highly nonlinear response. In the load transmission from structure to piping/equipment significant response amplifications and shifts to higher frequencies occurred.more » The performance of various pipe support configurations was evaluated. This latter effort was continued in the second series of tests (SHAM), in which an in-plant piping system was investigated at simulated seismic loads (generated by two servo-hydraulic actuators each capable of generating 40 tonnes of force), that exceeded design levels manifold and resulted in considerable pipe plastification and failure of some supports (snubbers). The evaluation of six different support configurations demonstrated that proper system design (for a given spectrum) rather than number of supports or system stiffness is essential to limiting pipe stresses. Pipe strains at loads exceeding the design level eightfold were still tolerable, indicating that pipe failure even under extreme seismic loads is unlikely inspite of multiple support failures. Conservatively, an excess capacity (margin) of at least four was estimated for the piping system, and the pipe damping was found to be 4%. Comparisons of linear and nonlinear computational results with measurements showed that analytical predictions have wide scatter and do not necessarily yield conservative responses, underpredicting, in particular, peak support forces.« less

Fuel processing in integrated micro-structured heat-exchanger reactors

NASA Astrophysics Data System (ADS)

Kolb, G.; Schürer, J.; Tiemann, D.; Wichert, M.; Zapf, R.; Hessel, V.; Löwe, H.

Micro-structured fuel processors are under development at IMM for different fuels such as methanol, ethanol, propane/butane (LPG), gasoline and diesel. The target application are mobile, portable and small scale stationary auxiliary power units (APU) based upon fuel cell technology. The key feature of the systems is an integrated plate heat-exchanger technology which allows for the thermal integration of several functions in a single device. Steam reforming may be coupled with catalytic combustion in separate flow paths of a heat-exchanger. Reactors and complete fuel processors are tested up to the size range of 5 kW power output of a corresponding fuel cell. On top of reactor and system prototyping and testing, catalyst coatings are under development at IMM for numerous reactions such as steam reforming of LPG, ethanol and methanol, catalytic combustion of LPG and methanol, and for CO clean-up reactions, namely water-gas shift, methanation and the preferential oxidation of carbon monoxide. These catalysts are investigated in specially developed testing reactors. In selected cases 1000 h stability testing is performed on catalyst coatings at weight hourly space velocities, which are sufficiently high to meet the demands of future fuel processing reactors.

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

Comparing mesophilic and thermophilic anaerobic digestion of chicken manure: Microbial community dynamics and process resilience

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

Niu, Qigui; Takemura, Yasuyuki; Kubota, Kengo

Highlights: • Microbial community dynamics and process functional resilience were investigated. • The threshold of TAN in mesophilic reactor was higher than the thermophilic reactor. • The recoverable archaeal community dynamic sustained the process resilience. • Methanosarcina was more sensitive than Methanoculleus on ammonia inhibition. • TAN and FA effects the dynamic of hydrolytic and acidogenic bacteria obviously. - Abstract: While methane fermentation is considered as the most successful bioenergy treatment for chicken manure, the relationship between operational performance and the dynamic transition of archaeal and bacterial communities remains poorly understood. Two continuous stirred-tank reactors were investigated under thermophilic andmore » mesophilic conditions feeding with 10%TS. The tolerance of thermophilic reactor on total ammonia nitrogen (TAN) was found to be 8000 mg/L with free ammonia (FA) 2000 mg/L compared to 16,000 mg/L (FA1500 mg/L) of mesophilic reactor. Biomethane production was 0.29 L/gV S{sub in} in the steady stage and decreased following TAN increase. After serious inhibition, the mesophilic reactor was recovered successfully by dilution and washing stratagem compared to the unrecoverable of thermophilic reactor. The relationship between the microbial community structure, the bioreactor performance and inhibitors such as TAN, FA, and volatile fatty acid was evaluated by canonical correspondence analysis. The performance of methanogenic activity and substrate removal efficiency were changed significantly correlating with the community evenness and phylogenetic structure. The resilient archaeal community was found even after serious inhibition in both reactors. Obvious dynamics of bacterial communities were observed in acidogenic and hydrolytic functional bacteria following TAN variation in the different stages.« less

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

Tan, Lizhen; Yang, Ying; Tyburska-Puschel, Beata

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 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 computational tools)more » is an important path to more efficient alloy development and process optimization. Ferritic-martensitic (FM) steels are important structural materials for nuclear reactors due to their advantages over other applicable materials like austenitic stainless steels, notably their resistance to void swelling, low thermal expansion coefficients, and higher thermal conductivity. However, traditional FM steels exhibit a noticeable yield strength reduction at elevated temperatures above ~500°C, which limits their applications in advanced nuclear reactors which target operating temperatures at 650°C or higher. Although oxide-dispersion-strengthened (ODS) ferritic steels have shown excellent high-temperature performance, their extremely high cost, limited size and fabricability of products, as well as the great difficulty with welding and joining, have limited or precluded their commercial applications. Zirconium has shown many benefits to Fe-base alloys such as grain refinement, improved phase stability, and reduced radiation-induced segregation. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of a new generation of Zr-bearing ferritic alloys to be fabricated using conventional steelmaking practices, which have excellent radiation resistance and enhanced high-temperature creep performance greater than Grade 91.« less

High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Test Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Kapernick, Richard

2007-01-01

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer. and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics and assess potential design improvements at relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design is developed: this is followed by engineering design, fabrication, and testing to validate the overall design process. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

Application of a novel type impinging streams reactor in solid-liquid enzyme reactions and modeling of residence time distribution using GDB model.

PubMed

Fatourehchi, Niloufar; Sohrabi, Morteza; Dabir, Bahram; Royaee, Sayed Javid; Haji Malayeri, Adel

2014-02-05

Solid-liquid enzyme reactions constitute important processes in biochemical industries. The isomerization of d-glucose to d-fructose, using the immobilized glucose isomerase (Sweetzyme T), as a typical example of solid-liquid catalyzed reactions has been carried out in one stage and multi-stage novel type of impinging streams reactors. Response surface methodology was applied to determine the effects of certain pertinent parameters of the process namely axial velocity (A), feed concentration (B), nozzles' flow rates (C) and enzyme loading (D) on the performance of the apparatus. The results obtained from the conversion of glucose in this reactor were much higher than those expected in conventional reactors, while residence time was decreased dramatically. Residence time distribution (RTD) in a one-stage impinging streams reactor was investigated using colored solution as the tracer. The results showed that the flow pattern in the reactor was close to that in a continuous stirred tank reactor (CSTR). Based on the analysis of flow region in the reactor, gamma distribution model with bypass (GDB) was applied to study the RTD of the reactor. The results indicated that RTD in the impinging streams reactor could be described by the latter model. Copyright © 2013 Elsevier Inc. All rights reserved.

Effect of phenol on the nitrogen removal performance and microbial community structure and composition of an anammox reactor.

PubMed

Pereira, Alyne Duarte; Leal, Cíntia Dutra; Dias, Marcela França; Etchebehere, Claudia; Chernicharo, Carlos Augusto L; de Araújo, Juliana Calabria

2014-08-01

The effects of phenol on the nitrogen removal performance of a sequencing batch reactor (SBR) with anammox activity and on the microbial community within the reactor were evaluated. A phenol concentration of 300 mg L(-1) reduced the ammonium-nitrogen removal efficiency of the SBR from 96.5% to 47%. The addition of phenol changed the microbial community structure and composition considerably, as shown by denaturing gradient gel electrophoresis and 454 pyrosequencing of 16S rRNA genes. Some phyla, such as Proteobacteria, Verrucomicrobia, and Firmicutes, increased in abundance, whereas others, such as Acidobacteria, Chloroflexi, Planctomycetes, GN04, WS3, and NKB19, decreased. The diversity of the anammox bacteria was also affected by phenol: sequences related to Candidatus Brocadia fulgida were no longer detected, whereas sequences related to Ca. Brocadia sp. 40 and Ca. Jettenia asiatica persisted. These results indicate that phenol adversely affects anammox metabolism and changes the bacterial community within the anammox reactor. Copyright © 2014 Elsevier Ltd. All rights reserved.

Three-dimensional time-dependent STAR reactor kinetics analyses coupled with RETRAN and MCPWR system response

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

Feltus, M.A.

1989-11-01

The operation of a nuclear power plant must be regularly supported by various reactor dynamics and thermal-hydraulic analyses, which may include final safety analysis report (FSAR) design-basis calculations, and conservative and best-estimate analyses. The development and improvement of computer codes and analysis methodologies provide many advantages, including the ability to evaluate the effect of modeling simplifications and assumptions made in previous reactor kinetics and thermal-hydraulic calculations. This paper describes the results of using the RETRAN, MCPWR, and STAR codes in a tandem, predictive-corrective manner for three pressurized water reactor (PWR) transients: (a) loss of feedwater (LOF) anticipated transient without scrammore » (ATWS), (b) station blackout ATWS, and (c) loss of total reactor coolant system (RCS) flow with a scram.« less

The necessity of nuclear reactors for targeted radionuclide therapies.

PubMed

Krijger, Gerard C; Ponsard, Bernard; Harfensteller, Mark; Wolterbeek, Hubert T; Nijsen, Johannes W F

2013-07-01

Nuclear medicine has been contributing towards personalized therapies. Nuclear reactors are required for the working horses of both diagnosis and treatment, i.e., Tc-99m and I-131. In fact, reactors will remain necessary to fulfill the demand for a variety of radionuclides and are essential in the expanding field of targeted radionuclide therapies for cancer. However, the main reactors involved in the global supply are ageing and expected to shut down before 2025. Therefore, the fields of (nuclear) medicine, nuclear industry and politics share a global responsibility, faced with the task to secure future access to suitable nuclear reactors. At the same time, alternative production routes should be industrialized. For this, a coordinating entity should be put into place. Copyright © 2013 Elsevier Ltd. All rights reserved.

Building on knowledge base of sodium cooled fast spectrum reactors to develop materials technology for fusion reactors

NASA Astrophysics Data System (ADS)

Raj, Baldev; Rao, K. Bhanu Sankara

2009-04-01

The alloys 316L(N) and Mod. 9Cr-1Mo steel are the major structural materials for fabrication of structural components in sodium cooled fast reactors (SFRs). Various factors influencing the mechanical behaviour of these alloys and different modes of deformation and failure in SFR systems, their analysis and the simulated tests performed on components for assessment of structural integrity and the applicability of RCC-MR code for the design and validation of components are highlighted. The procedures followed for optimal design of die and punch for the near net shape forming of petals of main vessel of 500 MWe prototype fast breeder reactor (PFBR); the safe temperature and strain rate domains established using dynamic materials model for forming of 316L(N) and 9Cr-1Mo steels components by various industrial processes are illustrated. Weldability problems associated with 316L(N) and Mo. 9Cr-1Mo are briefly discussed. The utilization of artificial neural network models for prediction of creep rupture life and delta-ferrite in austenitic stainless steel welds is described. The usage of non-destructive examination techniques in characterization of deformation, fracture and various microstructural features in SFR materials is briefly discussed. Most of the experience gained on SFR systems could be utilized in developing science and technology for fusion reactors. Summary of the current status of knowledge on various aspects of fission and fusion systems with emphasis on cross fertilization of research is presented.

LOGIC CIRCUIT

DOEpatents

Strong, G.H.; Faught, M.L.

1963-12-24

A device for safety rod counting in a nuclear reactor is described. A Wheatstone bridge circuit is adapted to prevent de-energizing the hopper coils of a ball backup system if safety rods, sufficient in total control effect, properly enter the reactor core to effect shut down. A plurality of resistances form one arm of the bridge, each resistance being associated with a particular safety rod and weighted in value according to the control effect of the particular safety rod. Switching means are used to switch each of the resistances in and out of the bridge circuit responsive to the presence of a particular safety rod in its effective position in the reactor core and responsive to the attainment of a predetermined velocity by a particular safety rod enroute to its effective position. The bridge is unbalanced in one direction during normal reactor operation prior to the generation of a scram signal and the switching means and resistances are adapted to unbalance the bridge in the opposite direction if the safety rods produce a predetermined amount of control effect in response to the scram signal. The bridge unbalance reversal is then utilized to prevent the actuation of the ball backup system, or, conversely, a failure of the safety rods to produce the predetermined effect produces no unbalance reversal and the ball backup system is actuated. (AEC)

Comparison of UWCC MOX fuel measurements to MCNP-REN calculations

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

Abhold, M.; Baker, M.; Jie, R.

1998-12-31

The development of neutron coincidence counting has greatly improved the accuracy and versatility of neutron-based techniques to assay fissile materials. Today, the shift register analyzer connected to either a passive or active neutron detector is widely used by both domestic and international safeguards organizations. The continued development of these techniques and detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model, as it is currently used, fails to accurately predict detector response in highly multiplying mediums such as mixed-oxide (MOX) lightmore » water reactor fuel assemblies. For this reason, efforts have been made to modify the currently used Monte Carlo codes and to develop new analytical methods so that this model is not required to predict detector response. The authors describe their efforts to modify a widely used Monte Carlo code for this purpose and also compare calculational results with experimental measurements.« less

Renewable Wood Pulp Paper Reactor with Hierarchical Micro/Nanopores for Continuous-Flow Nanocatalysis.

PubMed

Koga, Hirotaka; Namba, Naoko; Takahashi, Tsukasa; Nogi, Masaya; Nishina, Yuta

2017-06-22

Continuous-flow nanocatalysis based on metal nanoparticle catalyst-anchored flow reactors has recently provided an excellent platform for effective chemical manufacturing. However, there has been limited progress in porous structure design and recycling systems for metal nanoparticle-anchored flow reactors to create more efficient and sustainable catalytic processes. In this study, traditional paper is used for a highly efficient, recyclable, and even renewable flow reactor by tailoring the ultrastructures of wood pulp. The "paper reactor" offers hierarchically interconnected micro- and nanoscale pores, which can act as convective-flow and rapid-diffusion channels, respectively, for efficient access of reactants to metal nanoparticle catalysts. In continuous-flow, aqueous, room-temperature catalytic reduction of 4-nitrophenol to 4-aminophenol, a gold nanoparticle (AuNP)-anchored paper reactor with hierarchical micro/nanopores provided higher reaction efficiency than state-of-the-art AuNP-anchored flow reactors. Inspired by traditional paper materials, successful recycling and renewal of AuNP-anchored paper reactors were also demonstrated while high reaction efficiency was maintained. © 2017 The Authors. Published by Wiley-VCH Verlag GmbH & Co. KGaA.

Reactor application of an improved bundle divertor

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

Yang, T.F.; Ruck, G.W.; Lee, A.Y.

1978-11-01

A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supportedmore » by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW.« less

Applicability of 100kWe-class of space reactor power systems to NASA manned space station missions

NASA Technical Reports Server (NTRS)

Silverman, S. W.; Willenberg, H. J.; Robertson, C.

1985-01-01

An assessment is made of a manned space station operating with sufficiently high power demands to require a multihundred kilowatt range electrical power system. The nuclear reactor is a competitor for supplying this power level. Load levels were selected at 150kWe and 300kWe. Interactions among the reactor electrical power system, the manned space station, the space transportation system, and the mission were evaluated. The reactor shield and the conversion equipment were assumed to be in different positions with respect to the station; on board, tethered, and on a free flyer platform. Mission analyses showed that the free flyer concept resulted in unacceptable costs and technical problems. The tethered reactor providing power to an electrolyzer for regenerative fuel cells on the space station, results in a minimum weight shield and can be designed to release the reactor power section so that it moves to a high altitude orbit where the decay period is at least 300 years. Placing the reactor on the station, on a structural boom is an attractive design, but heavier than the long tethered reactor design because of the shield weight for manned activity near the reactor.

EFFECT OF MOISTURE ON ADSORPTION OF ELEMENTAL MERCURY BY ACTIVATED CARBON

EPA Science Inventory

The paper discusses experiments using activated carbon to capture elemental mercury (Hgo), and a bench-scale dixed-bed reactor and a flow reactor to determine the role of surface moisture in Hgo adsorption. Three activated-carbon samples, with different pore structure and ash co...

Reactivity Initiated Accident Simulation to Inform Transient Testing of Candidate Advanced Cladding

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

Brown, Nicholas R; Wysocki, Aaron J; Terrani, Kurt A

2016-01-01

Abstract. Advanced cladding materials with potentially enhanced accident tolerance will yield different light water reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to different cladding material properties and responses to the transient, and to some extent, reactor physics, thermal, and hydraulic characteristics. Some of the differences in reactors physics characteristics will be driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and others will be driven by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermalmore » hydraulic limits after transition from the current zirconium-based cladding to the advanced materials will also affect the transient response of the integral fuel. This paper leverages three-dimensional reactor core simulation capabilities to inform on appropriate experimental test conditions for candidate advanced cladding materials in a control rod ejection event. These test conditions are using three-dimensional nodal kinetics simulations of a reactivity initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon carbide based (SiC-SiC) cladding materials. The effort yields boundary conditions for experimental mechanical tests, specifically peak cladding strain during the power pulse following the rod ejection. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus reference zirconium cladding is predominantly due to differences in: (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores due to hardened (or softened) spectrum. This study shows minimal impact of SiC-based cladding configurations on the transient response versus reference zirconium-based cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. Therefore the FeCrAl-based cases have a more rapid fuel thermal expansion rate and the resultant pellet-cladding interaction occurs more rapidly.« less

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

NASA Astrophysics Data System (ADS)

Koshelev, A. S.; Kovshov, K. N.; Ovchinnikov, M. A.; Pikulina, G. N.; Sokolov, A. B.

2016-12-01

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

Materials technology for an advanced space power nuclear reactor concept: Program summary

NASA Technical Reports Server (NTRS)

Gluyas, R. E.; Watson, G. K.

1975-01-01

The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

Shield Design for Lunar Surface Applications

NASA Astrophysics Data System (ADS)

Johnson, Gregory A.

2006-01-01

A shielding concept for lunar surface applications of nuclear power is presented herein. The reactor, primary shield, reactor equipment and power generation module are placed in a cavity in the lunar surface. Support structure and heat rejection radiator panels are on the surface, outside the cavity. The reactor power of 1,320 kWt was sized to deliver 50 kWe from a thermoelectric power conversion subsystem. The dose rate on the surface is less than 0.6 mRem/hr at 100 meters from the reactor. Unoptimized shield mass is 1,020 kg which is much lighter than a comparable 4π shield weighing in at 17,000 kg.

Spectral structure of electron antineutrinos from nuclear reactors.

PubMed

Dwyer, D A; Langford, T J

2015-01-09

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

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

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

Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Kovshov, K. N.; Ovchinnikov, M. A.

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

NASA Astrophysics Data System (ADS)

Scarlat, Raluca O.; Peterson, Per F.

2014-01-01

The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

Strategy to identify the causes and to solve a sludge granulation problem in methanogenic reactors: application to a full-scale plant treating cheese wastewater.

PubMed

Macarie, Hervé; Esquivel, Maricela; Laguna, Acela; Baron, Olivier; El Mamouni, Rachid; Guiot, Serge R; Monroy, Oscar

2017-08-26

Granulation of biomass is at the basis of the operation of the most successful anaerobic systems (UASB, EGSB and IC reactors) applied worldwide for wastewater treatment. Despite of decades of studies of the biomass granulation process, it is still not fully understood and controlled. "Degranulation/lack of granulation" is a problem that occurs sometimes in anaerobic systems resulting often in heavy loss of biomass and poor treatment efficiencies or even complete reactor failure. Such a problem occurred in Mexico in two full-scale UASB reactors treating cheese wastewater. A close follow-up of the plant was performed to try to identify the factors responsible for the phenomenon. Basically, the list of possible causes to a granulation problem that were investigated can be classified amongst nutritional, i.e. related to wastewater composition (e.g. deficiency or excess of macronutrients or micronutrients, too high COD proportion due to proteins or volatile fatty acids, high ammonium, sulphate or fat concentrations), operational (excessive loading rate, sub- or over-optimal water upflow velocity) and structural (poor hydraulic design of the plant). Despite of an intensive search, the causes of the granulation problems could not be identified. The present case remains however an example of the strategy that must be followed to identify these causes and could be used as a guide for plant operators or consultants who are confronted with a similar situation independently of the type of wastewater. According to a large literature based on successful experiments at lab scale, an attempt to artificially granulate the industrial reactor biomass through the dosage of a cationic polymer was also tested but equally failed. Instead of promoting granulation, the dosage caused a heavy sludge flotation. This shows that the scaling of such a procedure from lab to real scale cannot be advised right away unless its operability at such a scale can be demonstrated.

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

Mohanty, Subhasish; Majumdar, Saurindranath

Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector.

TEMPEST code simulations of hydrogen distribution in reactor containment structures. Final report

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

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

The mass transport version of the TEMPEST computer code was used to simulate hydrogen distribution in geometric configurations relevant to reactor containment structures. Predicted results of Battelle-Frankfurt hydrogen distribution tests 1 to 6, and 12 are presented. Agreement between predictions and experimental data is good. Best agreement is obtained using the k-epsilon turbulence model in TEMPEST in flow cases where turbulent diffusion and stable stratification are dominant mechanisms affecting transport. The code's general analysis capabilities are summarized.

A ``NEW'' Solid-Core Reactor Fuel Form that Maximizes the Performance of Nuclear Thermal and Electric Rockets

NASA Astrophysics Data System (ADS)

Rom, Frank E.; Finnegan, Patrick M.

1994-07-01

The ``NEW'' solid-core fuel form is the old Vapor Transport (VT) fuel pin investigated at NASA about 30 years ago. It is simply a tube sealed at both ends partially filled with UO2. During operation the UO2 forms an annular layer on the inside of the tube by vaporization and condensation. This form is an ideal structure for overall strength and retention of fission products. All of the structural material lies between the fuel (including fission products) and the reactor coolant. The isothermal inside fuel surface temperature that results from the vaporization and condensation of fuel during operation eliminates hotspots, significantly increasing the design fuel pin surface temperature. For NTP, W-UO2 fuel pins yield higher operating temperatures than for other fuel forms, because W has about a ten-fold lower vaporization rate compared to any other known material. The use of perigee propulsion using W-UO2 fuel pins can result in a more than ten-fold reduction in reactor power. Lower reactor power, together with zero fission product release potential, and the simplicity of fabrication of VT fuel pins should greatly simplify and reduce the cost of development of NTP. For NEP, VT fuel pins can increase fast neutron spectrum reactor life with no fission product release. Thermal spectrum NEP reactors using W184 or Mo VT fuel pins, with only small amounts of high neutron absorbing additives, offer benefits because of much lower fissionable fuel requirements. The VT fuel pin has application to commercial power reactors with similar benefits.

Reactor Dosimetry State of the Art 2008

NASA Astrophysics Data System (ADS)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G. Williams, A. P. Ribaric and T. Schnauber. Agile high-fidelity MCNP model development techniques for rapid mechanical design iteration / J. A. Kulesza.Extension of Raptor-M3G to r-8-z geometry for use in reactor dosimetry applications / M. A. Hunter, G. Longoni and S. L. Anderson. In vessel exposure distributions evaluated with MCNP5 for Atucha II / J. M. Longhino, H. Blaumann and G. Zamonsky. Atucha I nuclear power plant azimutal ex-vessel flux profile evaluation / J. M. Longhino ... [et al.]. UFTR thermal column characterization and redesign for maximized thermal flux / C. Polit and A. Haghighat. Activation counter using liquid light-guide for dosimetry of neutron burst / M. Hayashi ... [et al.]. Control rod reactivity curves for the annular core research reactor / K. R. DePriest ... [et al.]. Specification of irradiation conditions in VVER-440 surveillance positions / V. Kochkin ... [et al.]. Simulations of Mg-Ar ionisation and TE-TE ionisation chambers with MCNPX in a straightforward gamma and beta irradiation field / S. Nievaart ... [et al.]. The change of austenitic stainless steel elements content in the inner parts of VVER-440 reactor during operation / V. Smutný, J. Hep and P. Novosad. Fast neutron environmental spectrometry using disk activation / G. Lövestam ... [et al.]. Optimization of the neutron activation detector location scheme for VVER-lOOO ex-vessel dosimetry / V. N. Bukanov ... [et al.]. Irradiation conditions for surveillance specimens located into plane containers installed in the WWER-lOOO reactor of unit 2 of the South-Ukrainian NPP / O. V. Grytsenko. V. N. Bukanov and S. M. Pugach. Conformity between LRO mock-ups and VVERS NPP RPV neutron flux attenuation / S. Belousov. Kr. Ilieva and D. Kirilova. FLUOLE: a new relevant experiment for PWR pressure vessel surveillance / D. Beretz ... [et al.]. Transport of neutrons and photons through the iron and water layers / M. J. Kost'ál ... [et al.]. Condition evaluation of spent nuclear fuel assemblies from the first-generation nuclear-powered submarines by gamma scanning / A. F. Usatyi. L. A. Serdyukova and B. S. Stepennov -- Oral session 3: Power plant surveillance. Upgraded neutron dosimetry procedure for VVER-440 surveillance specimens / V. Kochkin ... [et al.]. Neutron dosimetry on the full-core first generation VVER-440 aimed to reactor support structure load evaluation / P. Borodkin ... [et al.]. Ex-vessel neutron dosimetry programs for PWRs in Korea / C. S. Yoo. B. C. Kim and C. C. Kim. Comparison of irradiation conditions of VVER-1000 reactor pressure vessel and surveillance specimens for various core loadings / V. N. Bukanov ... [et al.]. Re-evaluation of dosimetry in the new surveillance program for the Loviisa 1 VVER-440 reactor / T. Serén -- Oral session 4: Benchmarks, intercomparisons and adjustment methods. Determination of the neutron parameter's uncertainties using the stochastic methods of uncertainty propagation and analysis / G. Grégoire ... [et al.].Covariance matrices for calculated neutron spectra and measured dosimeter responses / J. G. Williams ... [et al.]. The role of dosimetry at the high flux reactor / S. C. van der Marek ... [et al.]. Calibration of a manganese bath relative to Cf-252 nu-bar / D. M. Gilliam, A. T. Yue and M. Scott Dewey. Major upgrade of the reactor dosimetry interpretation methodology used at the CEA: general principle / C. Destouches ... [et al.] -- Oral session 5: power plant surveillance. The role of ex-vessel neutron dosimetry in reactor vessel surveillance in South Korea / B.-C. Kim ... [et al.]. Spanish RPV surveillance programmes: lessons learned and current activities / A. Ballesteros and X. Jardí. Atucha I nuclear power plant extended dosimetry and assessment / H. Blaumann ... [et al.]. Monitoring of radiation load of pressure vessels of Russian VVER in compliance with license amendments / G. Borodkin ... [et al.] -- Poster session 2: Test reactors, accelerators and advanced systems; cross sections, nuclear data, damage correlations. Two-dimensional mapping of the calculated fission power for the full-size fuel plate experiment irradiated in the advanced test reactor / G. S. Chang and M. A. Lillo. The radiation safety information computational center: a resource for reactor dosimetry software and nuclear data / B. L. Kirk. Irradiated xenon isotopic ratio measurement for failed fuel detection and location in fast reactor / C. Ito, T. Iguchi and H. Harano. Characterization of dosimetry of the BMRR horizontal thimble tubes and broad beam facility / J.-P. Hu, R. N. Reciniello and N. E. Holden. 2007 nuclear data review / N. E. Holden. Further dosimetry studies at the Rhode Island nuclear science / R. N. Reciniello ... [et al.]. Characterization of neutron fields in the experimental fast reactor Joyo MK-III core / S. Maeda ... [et al.]. Measuring [symbol]Li(n, t) and [symbol]B(n, [symbol]) cross sections using the NIST alpha-gamma apparatus / M. S. Dewey ... [et al.]. Improvement of neutron/gamma field evaluation for restart of JMTR / Y. Nagao ... [et al.]. Monitoring of the irradiated neutron fluence in the neutron transmutation doping process of HANARO / M.-S. Kim and S.-J. Park.Training reactor VR-l neutron spectrum determination / M. Vins, A. Kolros and K. Katovsky. Differential cross sections for gamma-ray production by 14 MeV neutrons on iron and bismuth / V. M. Bondar ... [et al.]. The measurements of the differential elastic neutron cross-sections of carbon for energies from 2 to 133 ke V / O. Gritzay ... [et al.]. Determination of neutron spectrum by the dosimetry foil method up to 35 Me V / S. P. Simakov ... [et al.]. Extension of the BGL broad group cross section library / D. Kirilova, S. Belousov and Kr. Ilieva. Measurements of neutron capture cross-section for tantalum at the neutron filtered beams / O. Gritzayand V. Libman. Measurements of microscopic data at GELINA in support of dosimetry / S. Kopecky ... [et al.]. Nuclide guide and international chart of nuclides - 2008 / T. Golashvili -- Oral session 6: Test reactors, accelerators and advanced systems. Neutronic analyses in support of the HFIR beamline modifications and lifetime extension / I. Remec and E. D. Blakeman. Characterization of neutron test facilities at Sandia National Laboratories / D. W. Vehar ... [et al.]. LYRA irradiation experiments: neutron metrology and dosimetry / B. Acosta and L. Debarberis. Calculated neutron and gamma-ray spectra across the prismatic very high temperature reactor core / J. W. Sterbentz. Enhancement of irradiation capability of the experimental fast reactor joyo / S. Maeda ... [et al.]. Neutron spectrum analyses by foil activation method for high-energy proton beams / C. H. Pyeon ... [et al.] -- Oral session 7: Cross sections, nuclear data, damage correlations. Investigation of new reaction cross-section evaluations in order to update and extend the IRDF-2002 reactor dosimetry library / É. M. Zsolnay, H. J. Nolthenius and A. L. Nichols. A novel approach towards DPA calculations / A. Hogenbirk and D. F. Da Cruz. A new ENDFIB-VII.O based multigroup cross-section library for reactor dosimetry / F. A. Alpan and S. L. Anderson. Activities at the NEA for dosimetry applications / H. Henriksson and I. Kodeli. Validation and verification of covariance data from dosimetry reaction cross-section evaluations / S. Badikov. Status of the neutron cross section standards / A. D. Carlson -- Oral session 8: transport calculations. A dosimetry assessment for the core restraint of an advanced gas cooled reactor / D. A. Thornton ... [et al.]. Neutron dosimetry study in the region of the support structure of a VVER-1000 type reactor / G. Borodkin ... [et al.]. SNS moderator poison design and experiment validation of the moderator performance / W. Lu ... [et al.]. Analysis of OSIRIS in-core surveillance dosimetry for GONDOLE steel irradiation program by using TRIPOLI-4 Monte Carlo code / Y. K. Lee and F. Malouch.Reactor dosimetry applications using RAPTOR-M3G: a new parallel 3-D radiation transport code / G. Longoni and S. L. Anderson.

The Fukushima Nuclear Disaster and the U.S. Customs and Border Protection Response

NASA Astrophysics Data System (ADS)

McCormick, Kathy

2013-10-01

On 3/11/11, the reactors at the Fukushima Nuclear Plant in Japan were damaged by a magnitude 9.0 earthquake. Of the six reactors at the site, three were in operation prior to the event, and were automatically shut-down during the earthquake. Emergency cooling systems came online and were subsequently destroyed by a tsunami generated by the earthquake. For the operating reactors, all the reactor cores were exposed, resulting in overheating and the release of steam and hydrogen gas to the containment vessels, several of which subsequently exploded, releasing radioactivity into the atmosphere. The cores of the operating reactors melted down, and radioactive water was released to the ocean in cooling efforts. The primary radiation concerns in the United States from the disaster were radioactive plumes driven by westerly winds and contaminated commercial products and travelers. In the United States, one of the primary governmental organizations to respond to the disaster was U.S. Customs and Border Protection (CBP), which has responsibility to oversee the safety and security of cargo and travelers entering the United States. This talk will describe the various types of radioactive commodities and events encountered by CBP in the U.S. from the Fukushima disaster. Thanks to the CBP Teleforensics Center for their assistance with this presentation.

Update on reactors and reactor instruments in Asia

NASA Astrophysics Data System (ADS)

Rao, K. R.

1991-10-01

The 1980s have seen the commissioning of several medium flux (∼10 14 neutrons/cm 2s) research reactors in Asia. The reactors are based on indigenous design and development in India and China. At Dhruva reactor (India), a variety of neutron spectrometers have been established that have provided useful data related to the structure of high- Tc materials, phonon density of states, magnetic moment distributions and micellar aggregation during the last couple of years. Polarised neutron analysis, neutron interferometry and neutron spin echo methods are some of the new techniques under development. The spectrometers and associated automaton, detectors and neutron guides have all been indigenously developed. This paper summarises the developments and on-going activities in Bangladesh, China, India, Indonesia, Korea, Malaysia, the Philippines and Thailand.

Long lifetime fast spectrum reactor for lunar surface power system

NASA Astrophysics Data System (ADS)

Kambe, Mitsuru

1993-01-01

In the framework of innovative reactor research activities, a conceptual design study of fast spectrum reactor and primary system for 800 kWe lunar surface power system to be combined with potassium Rankine cycle power conversion has been conducted to meet the power requirements of the lunar base activities in the next century. The reactor subsystem is characterized by RAPID (Refueling by All Pins Integrated Design) concept to enhance inherent safety and to enable quick and simplifed refueling in every 10 years. RAPID concept affords power plant design lifetime of up to 30 years. Integrity of the reactor structure and replacement of failed primary circuits are also discussed. Substantial reduction in per-kWh cost on considering launch, emplacement, and final disposition can be expected by a long system lifetime.

Structural modifications induced by ion irradiation and temperature in boron carbide B4C

NASA Astrophysics Data System (ADS)

Victor, G.; Pipon, Y.; Bérerd, N.; Toulhoat, N.; Moncoffre, N.; Djourelov, N.; Miro, S.; Baillet, J.; Pradeilles, N.; Rapaud, O.; Maître, A.; Gosset, D.

2015-12-01

Already used as neutron absorber in the current French nuclear reactors, boron carbide (B4C) is also considered in the future Sodium Fast Reactors of the next generation (Gen IV). Due to severe irradiation conditions occurring in these reactors, it is of primary importance that this material presents a high structural resistance under irradiation, both in the ballistic and electronic damage regimes. Previous works have shown an important structural resistance of boron carbide even at high neutron fluences. Nevertheless, the structural modification mechanisms due to irradiation are not well understood. Therefore the aim of this paper is to study structural modifications induced in B4C samples in different damage regimes. The boron carbide pellets were shaped and sintered by using spark plasma sintering method. They were then irradiated in several conditions at room temperature or 800 °C, either by favoring the creation of ballistic damage (between 1 and 3 dpa), or by favoring the electronic excitations using 100 MeV swift iodine ions (Se ≈ 15 keV/nm). Ex situ micro-Raman spectroscopy and Doppler broadening of annihilation radiation technique with variable energy slow positrons were coupled to follow the evolution of the B4C structure under irradiation.

Generation of the V4.2m5 and AMPX and MPACT 51 and 252-Group Libraries with ENDF/B-VII.0 and VII.1

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

Kim, Kang Seog

The evaluated nuclear data file (ENDF)/B-7.0 v4.1m3 MPACT 47-group library has been used as a main library for the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronics simulator in simulating pressurized water reactor (PWR) problems. Recent analysis for the high void boiling water reactor (BWR) fuels and burnt fuels indicates that the 47-group library introduces relatively large reactivity bias. Since the 47- group structure does not match with the SCALE 6.2 252-group boundaries, the CASL Virtual Environment for Reactor Applications Core Simulator (VERA-CS) MPACT library must be maintained independently, which causes quality assurance concerns. In order to addressmore » this issue, a new 51-group structure has been proposed based on the MPACT 47- g and SCALE 252-g structures. In addition, the new CASL library will include a 19-group structure for gamma production and interaction cross section data based on the SCALE 19- group structure. New AMPX and MPACT 51-group libraries have been developed with the ENDF/B-7.0 and 7.1 evaluated nuclear data. The 19-group gamma data also have been generated for future use, but they are only available on the AMPX 51-g library. In addition, ENDF/B-7.0 and 7.1 MPACT 252-g libraries have been generated for verification purposes. Various benchmark calculations have been performed to verify and validate the newly developed libraries.« less

Corrosion and stress corrosion cracking in supercritical water

NASA Astrophysics Data System (ADS)

Was, G. S.; Ampornrat, P.; Gupta, G.; Teysseyre, S.; West, E. A.; Allen, T. R.; Sridharan, K.; Tan, L.; Chen, Y.; Ren, X.; Pister, C.

2007-09-01

Supercritical water (SCW) has attracted increasing attention since SCW boiler power plants were implemented to increase the efficiency of fossil-based power plants. The SCW reactor (SCWR) design has been selected as one of the Generation IV reactor concepts because of its higher thermal efficiency and plant simplification as compared to current light water reactors (LWRs). Reactor operating conditions call for a core coolant temperature between 280 °C and 620 °C at a pressure of 25 MPa and maximum expected neutron damage levels to any replaceable or permanent core component of 15 dpa (thermal reactor design) and 100 dpa (fast reactor design). Irradiation-induced changes in microstructure (swelling, radiation-induced segregation (RIS), hardening, phase stability) and mechanical properties (strength, thermal and irradiation-induced creep, fatigue) are also major concerns. Throughout the core, corrosion, stress corrosion cracking, and the effect of irradiation on these degradation modes are critical issues. This paper reviews the current understanding of the response of candidate materials for SCWR systems, focusing on the corrosion and stress corrosion cracking response, and highlights the design trade-offs associated with certain alloy systems. Ferritic-martensitic steels generally have the best resistance to stress corrosion cracking, but suffer from the worst oxidation. Austenitic stainless steels and Ni-base alloys have better oxidation resistance but are more susceptible to stress corrosion cracking. The promise of grain boundary engineering and surface modification in addressing corrosion and stress corrosion cracking performance is discussed.

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.

FOEHN: The critical experiment for the Franco-German High Flux Reactor

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

Scharmer, K.; Eckert, H. G.

1991-01-01

A critical experiment for the Franco-German High Flux Reactor was carried out in the French reactor EOLE (CEN Cadarache). The purpose of the experiment was to check the calculation methods in a realistic geometry and to measure effects that can only be calculated imprecisely (e.g. beam hole effects). The structure of the experiment and the measurement and calculation methods are described. A detailed comparison between theoretical and experimental results was performed. 30 refs., 105 figs.

11. Photocopy of drawing, February 1958. WATERTOWN ARSENAL REACTOR, SHIELD ...

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

11. Photocopy of drawing, February 1958. WATERTOWN ARSENAL REACTOR, SHIELD STRUCTURE, SECTIONS, AND PLANS. Bendix Aviation Corporation; and Giffels & Vallet, Inc., L. Rosetti, Associated Architects and Engineers, Detroit, Michigan. Drawing Number 53-198. (Original: AMTL Engineering Division, Watertown). - Watertown Arsenal, Building No. 100, Wooley Avenue, Watertown, Middlesex County, MA

10 CFR 52.137 - Contents of applications; technical information.

Code of Federal Regulations, 2010 CFR

2010-01-01

... limits on its operation, and presents a safety analysis of the structures, systems, and components and of... products. The description shall be sufficient to permit understanding of the system designs and their relationship to the safety evaluations. Items such as the reactor core, reactor coolant system, instrumentation...

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

Jaffke, Patrick John

This acts as a short note on the effects of varying the value of the endpoints of the thermal, epithermal, and fast flux groups. As expected, varying these endpoints can alter the value of the cross-section for a given nuclide. This effect is quantified in this note for an important nuclide in reactor simulations, 238U. Uranium-238 is responsible for the production of Plutonium in most reactors, making it critical to understand all of the 238U capture modes leading to Plutonium. We explicitly quantify the reaction rates for 238U that are altered when we use a given research reactor fluxmore » and vary the endpoint definitions of said flux as well as the reactor position.« less

REACTOR CONTROL DEVICE

DOEpatents

Graham, R.H.

1962-09-01

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

CONTROL ROD ROTATING MECHANISM

DOEpatents

Baumgarten, A.; Karalis, A.J.

1961-11-28

A threaded rotatable shaft is provided which rotates in response to linear movement of a nut, the shaft being surrounded by a pair of bellows members connected to either side of the nut to effectively seal the reactor from leakage and also to store up energy to shut down the reactor in the event of a power failure. (AEC)

78 FR 71673 - Agency Information Collection Activities: Submission for the Office of Management and Budget (OMB...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-29

..., power reactors under construction, research and test reactors, agreement states, non-agreement states, as well as users of byproduct material (e.g. departments of health, medical centers, steel mills, well loggers, and radiographers.) 7. An estimate of the number of annual responses: 339. [[Page 71674...

Tailoring the response of Autonomous Reactivity Control (ARC) systems

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

Qvist, Staffan A.; Hellesen, Carl; Gradecka, Malwina

The Autonomous Reactivity Control (ARC) system was developed to ensure inherent safety of Generation IV reactors while having a minimal impact on reactor performance and economic viability. In this study we present the transient response of fast reactor cores to postulated accident scenarios with and without ARC systems installed. Using a combination of analytical methods and numerical simulation, the principles of ARC system design that assure stability and avoids oscillatory behavior have been identified. A comprehensive transient analysis study for ARC-equipped cores, including a series of Unprotected Loss of Flow (ULOF) and Unprotected Loss of Heat Sink (ULOHS) simulations, weremore » performed for Argonne National Laboratory (ANL) Advanced Burner Reactor (ABR) designs. With carefully designed ARC-systems installed in the fuel assemblies, the cores exhibit a smooth non-oscillatory transition to stabilization at acceptable temperatures following all postulated transients. To avoid oscillations in power and temperature, the reactivity introduced per degree of temperature change in the ARC system needs to be kept below a certain threshold the value of which is system dependent, the temperature span of actuation needs to be as large as possible.« less

Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors. Final Report

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

Hawari, Ayman; Ougouag, Abderrafi

2014-07-08

This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermalization is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can bemore » easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.« less

3D printing in chemical engineering and catalytic technology: structured catalysts, mixers and reactors.

PubMed

Parra-Cabrera, Cesar; Achille, Clement; Kuhn, Simon; Ameloot, Rob

2018-01-02

Computer-aided fabrication technologies combined with simulation and data processing approaches are changing our way of manufacturing and designing functional objects. Also in the field of catalytic technology and chemical engineering the impact of additive manufacturing, also referred to as 3D printing, is steadily increasing thanks to a rapidly decreasing equipment threshold. Although still in an early stage, the rapid and seamless transition between digital data and physical objects enabled by these fabrication tools will benefit both research and manufacture of reactors and structured catalysts. Additive manufacturing closes the gap between theory and experiment, by enabling accurate fabrication of geometries optimized through computational fluid dynamics and the experimental evaluation of their properties. This review highlights the research using 3D printing and computational modeling as digital tools for the design and fabrication of reactors and structured catalysts. The goal of this contribution is to stimulate interactions at the crossroads of chemistry and materials science on the one hand and digital fabrication and computational modeling on the other.

Postirradiation thermocyclic loading of ferritic-martensitic structural materials

NASA Astrophysics Data System (ADS)

Belyaeva, L.; Orychtchenko, A.; Petersen, C.; Rybin, V.

Thermonuclear fusion reactors of the Tokamak-type will be unique power engineering plants to operate in thermocyclic mode only. Ferritic-martensitic stainless steels are prime candidate structural materials for test blankets of the ITER fusion reactor. Beyond the radiation damage, thermomechanical cyclic loading is considered as the most detrimental lifetime limiting phenomenon for the above structure. With a Russian and a German facility for thermal fatigue testing of neutron irradiated materials a cooperation has been undertaken. Ampule devices to irradiate specimens for postirradiation thermal fatigue tests have been developed by the Russian partner. The irradiation of these ampule devices loaded with specimens of ferritic-martensitic steels, like the European MANET-II, the Russian 05K12N2M and the Japanese Low Activation Material F82H-mod, in a WWR-M-type reactor just started. A description of the irradiation facility, the qualification of the ampule device and the modification of the German thermal fatigue facility will be presented.

The soret effect and its implications for fusion reactors

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.

1985-03-01

Tritium permeation through and retention in fusion reactor structures may be strongly influenced by the heat load carried by the structures through the Soret effect. After a short discussion suggestive of a heuristic model for predicting the associated energy and the heat of transport, data from several experiments are analyzed to show that the simplistic model works reasonably well with endothermic materials such as Fe and Ni, but is less successful with hydride formers. The implications of the model for tritium permeation and retention are discussed, and sample calculations are presented to illustrate the importance of properly accounting for the Soret effect in predicting tritium permeation and retention in fusion reactor structures. Neglecting the Soret effect may result in order of magnitude errors in estimating permeation and retention, while accounting for temperature sensitivity in the heat of transport will result in less significant corrections. An Appendix summarizes the development of transport equations from non-equilibrium thermodynamics to clarify the relationships between the various transport parameters involved.

Steady performance of a zero valent iron packed anaerobic reactor for azo dye wastewater treatment under variable influent quality.

PubMed

Zhang, Yaobin; Liu, Yiwen; Jing, Yanwen; Zhao, Zhiqiang; Quan, Xie

2012-01-01

Zero valent iron (ZVI) is expected to help create an enhanced anaerobic environment that might improve the performance of anaerobic treatment. Based on this idea, a novel ZVI packed upflow anaerobic sludge blanket (ZVI-UASB) reactor was developed to treat azo dye wastewater with variable influent quality. The results showed that the reactor was less influenced by increases of Reactive Brilliant Red X-3B concentration from 50 to 1000 mg/L and chemical oxygen demand (COD) from 1000 to 7000 mg/L in the feed than a reference UASB reactor without the ZVI. The ZVI decreased oxidation-reduction potential in the reactor by about 80 mV. Iron ion dissolution from the ZVI could buffer acidity in the reactor, the amount of which was related to the COD concentration. Fluorescence in situ hybridization test showed the abundance of methanogens in the sludge of the ZVI-UASB reactor was significantly greater than that of the reference one. Denaturing gradient gel electrophoresis showed that the ZVI increased the diversity of microbial strains responsible for high efficiency.

Ammonia-oxidizing microbial communities in reactors with efficient nitrification at low-dissolved oxygen

PubMed Central

Fitzgerald, Colin M.; Camejo, Pamela; Oshlag, J. Zachary; Noguera, Daniel R.

2015-01-01

Ammonia-oxidizing microbial communities involved in ammonia oxidation under low dissolved oxygen (DO) conditions (<0.3 mg/L) were investigated using chemostat reactors. One lab-scale reactor (NS_LowDO) was seeded with sludge from a full-scale wastewater treatment plant (WWTP) not adapted to low-DO nitrification, while a second reactor (JP_LowDO) was seeded with sludge from a full-scale WWTP already achieving low-DO nitrifiaction. The experimental evidence from quantitative PCR, rDNA tag pyrosequencing, and fluorescence in situ hybridization (FISH) suggested that ammonia-oxidizing bacteria (AOB) in the Nitrosomonas genus were responsible for low-DO nitrification in the NS_LowDO reactor, whereas in the JP_LowDO reactor nitrification was not associated with any known ammonia-oxidizing prokaryote. Neither reactor had a significant population of ammonia-oxidizing archaea (AOA) or anaerobic ammonium oxidation (anammox) organisms. Organisms isolated from JP_LowDO were capable of autotrophic and heterotrophic ammonia utilization, albeit without stoichiometric accumulation of nitrite or nitrate. Based on the experimental evidence we propose that Pseudomonas, Xanthomonadaceae, Rhodococcus, and Sphingomonas are involved in nitrification under low-DO conditions. PMID:25506762

Nuclear Reactor Physics

NASA Astrophysics Data System (ADS)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

THE EXPERIENCE IN THE UNITED STATES WITH REACTOR OPERATION AND REACTOR SAFEGUARDS

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

McCullough, C.R.

1958-10-31

Reactors are operating or planned at locations in the United States in cities, near cities, and at remote locations. There is a general pattern that the higher power reactors are not in, but fairly uear cities, and the testing reactors for more hazardous experiments are at remote locations. A great deal has been done on the theoretical and experimental study of importunt features of reactor design. The metal-water reaction is still a theoretical possibility but tests of fuel element burnout under conditions approaching reactor operation gave no reaction. It appears that nucleate boiling does not necessarily result in steam blanketingmore » and fuel melting. Much attention is being given to the calculation of core kinetics but it is being found that temperature, power, and void coefficients cannot be calculated with accuracy and experiments are required. Some surprises are found giving positive localized void coefficients. Possible oscillatory behavior of reactors is being given careful study. No dangerous oscillations have been found in operating reactors but osciliations hare appeared in experimeats. The design of control and safety systems varies wvith different constructors. The relation of control to the kinetic behavior of the reactor is being studied. The importance of sensing element locations in order to know actual local reactor power level is being recognized. The time constants of instrumentation as related to reactor kinetics are being studied. Pressure vessels for reactors are being designed and manufactured. Many of these are beyond any previous experience. The stress problem is being given careful study. The effect of radiation is being studied experimentally. The stress problems of piping and pressure vessels is a difficult design problem being met successfully in reactor plants. The proper organization and procedure for operation of reactors is being evolved for resourch, testing, and power reactors. The importance of written standards and instructions for both normal and abnormal operating conditions is recogmized. Corfinement of radioactive materials either by tight steel shells, tight buildings, or semi-tight structures vented through filters is considered necessary in the United States. A discussion will be given of specifications, construction, and testing of these structures. The need for emergency plans has been stressed by recent experiences in radioactive releases. The problems of such plans to cover all grades of accidents will be discussed. The theoretical consequences of releases of radioactive materials have been studied and these results will be compared with actual experience. The problem of exposures from normal and abnormal operetion of reactors is a problem of desiga and operation on one hand and the amount of damage to be expected on the other. The safeguard problem is closely related to the acceptable doses of radiouctivity which the ICRP recommend. The future of atomic energy depends upon adequate safeguards and economical design and operation. Accepted criteria are required to guide designers as to the proper balance of caution and boldness. (auth)« less

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

Nygaard, E. T.; Williams, M. M. R.; Angelo, P. L.

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

Performance and microbial communities of Mn(II)-based autotrophic denitrification in a Moving Bed Biofilm Reactor (MBBR).

PubMed

Su, Jun Feng; Luo, Xian Xin; Wei, Li; Ma, Fang; Zheng, Sheng Chen; Shao, Si Cheng

2016-07-01

In this study, Mn(II) as electron donor was tested for the effects on denitrification in the MBBR under the conditions of initial nitrate concentration (10mgL(-1), 30mgL(-1), 50mgL(-1)), pH (5, 6, 7) and hydraulic retention time (HRT) (4h, 8h, 12h) which conducted by response surface methodology (RSM), the results demonstrated that the highest nitrate removal efficiency was occurred under the conditions of initial nitrate concentration of 47.64mgL(-1), HRT of 11.96h and pH 5.21. Analysis of SEM and flow cytometry suggested that microorganisms were immobilized on the Yu Long plastic carrier media successfully before the reactor began to operate. Furthermore, high-throughput sequencing was employed to characterize and compare the community compositions and structures of MBBR under the optimum conditions, the results showed that Pseudomonas sp. SZF15 was the dominant contributor for effective removal of nitrate in the MBBR. Copyright © 2016 Elsevier Ltd. All rights reserved.

Counteracting foaming caused by lipids or proteins in biogas reactors using rapeseed oil or oleic acid as antifoaming agents.

PubMed

Kougias, P G; Boe, K; Einarsdottir, E S; Angelidaki, I

2015-08-01

Foaming is one of the major operational problems in biogas plants, and dealing with foaming incidents is still based on empirical practices. Various types of antifoams are used arbitrarily to combat foaming in biogas plants, but without any scientific support this action can lead to serious deterioration of the methanogenic process. Many commercial antifoams are derivatives of fatty acids or oils. However, it is well known that lipids can induce foaming in manure based biogas plants. This study aimed to elucidate the effect of rapeseed oil and oleic acid on foam reduction and process performance in biogas reactors fed with protein or lipid rich substrates. The results showed that both antifoams efficiently suppressed foaming. Moreover rapeseed oil resulted in stimulation of the biogas production. Finally, it was reckoned that the chemical structure of lipids, and more specifically their carboxylic ends, is responsible for their foam promoting or foam counteracting behaviour. Thus, it was concluded that the fatty acids and oils could suppress foaming, while salt of fatty acids could generate foam. Copyright © 2015 Elsevier Ltd. All rights reserved.

Effect of torrefaction on structure and fast pyrolysis behavior of corncobs.

PubMed

Zheng, Anqing; Zhao, Zengli; Chang, Sheng; Huang, Zhen; Wang, Xiaobo; He, Fang; Li, Haibin

2013-01-01

Pretreatment of corncobs using torrefaction was conducted in an auger reactor at 250-300 °C and residence times of 10-60 min. The torrefied corncobs were fast pyrolyzed in a bubbling fluidized bed reactor at 470 °C to obtain high-quality bio-oil. The heating value and pH of the bio-oil improved when the torrefaction as pretreatment was applied; however, increasing bio-oil yield penalties were observed with increasing torrefaction severity. Fourier transform infrared Spectroscopy (FTIR) and quantitative solid (13)C nuclear magnetic resonance spectrometry (NMR) analysis of torrefied corncobs showed that the devolatilization, crosslinking and charring of corncobs during torrefaction could be responsible for the bio-oil yield penalties. Gas chromatography-mass spectrometry (GC-MS) analysis showed that the acetic acid and furfural contents of the bio-oil decreased with torrefaction temperature or residence time. The results showed that torrefaction is an effective method of pretreatment for improving bio-oil quality if the crosslinking and charring of biomass can be restricted. Copyright © 2012 Elsevier Ltd. All rights reserved.

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

Mendez Cruz, Carmen Margarita; Rochau, Gary E.; Middleton, Bobby

Sandia National Laboratories and General Atomics are pleased to respond to the Advanced Research Projects Agency-Energy (ARPA-e)’s request for information on innovative developments that may overcome various current reactor-technology limitations. The RFI is particularly interested in innovations that enable ultra-safe and secure modular nuclear energy systems. Our response addresses the specific features for reactor designs called out in the RFI, including a brief assessment of the current state of the technologies that would enable each feature and the methods by which they could be best incorporated into a reactor design.

Beta ray flux measuring device

DOEpatents

Impink, Jr., Albert J.; Goldstein, Norman P.

1990-01-01

A beta ray flux measuring device in an activated member in-core instrumentation system for pressurized water reactors. The device includes collector rings positioned about an axis in the reactor's pressure boundary. Activated members such as hydroballs are positioned within respective ones of the collector rings. A response characteristic such as the current from or charge on a collector ring indicates the beta ray flux from the corresponding hydroball and is therefore a measure of the relative nuclear power level in the region of the reactor core corresponding to the specific exposed hydroball within the collector ring.

Short- and long-term responses to molybdenum-99 shortages in nuclear medicine.

PubMed

Ballinger, J R

2010-11-01

Most nuclear medicine studies use (99)Tc(m), which is the decay product of (99)Mo. The world supply of (99)Mo comes from only five nuclear research reactors and availability has been much reduced in recent times owing to problems at the largest reactors. In the short-term there are limited actions that can be taken owing to capacity issues on alternative imaging modalities. In the long-term, stability of (99)Mo supply will rely on a combination of replacing conventional reactors and developing new technologies.

Short- and long-term responses to molybdenum-99 shortages in nuclear medicine

PubMed Central

Ballinger, J R

2010-01-01

Most nuclear medicine studies use 99Tcm, which is the decay product of 99Mo. The world supply of 99Mo comes from only five nuclear research reactors and availability has been much reduced in recent times owing to problems at the largest reactors. In the short-term there are limited actions that can be taken owing to capacity issues on alternative imaging modalities. In the long-term, stability of 99Mo supply will rely on a combination of replacing conventional reactors and developing new technologies. PMID:20965898

Balancing Near-Field Enhancement, Absorption, and Scattering for Effective Antenna-Reactor Plasmonic Photocatalysis.

PubMed

Li, Kun; Hogan, Nathaniel J; Kale, Matthew J; Halas, Naomi J; Nordlander, Peter; Christopher, Phillip

2017-06-14

Efficient photocatalysis requires multifunctional materials that absorb photons and generate energetic charge carriers at catalytic active sites to facilitate a desired chemical reaction. Antenna-reactor complexes are an emerging multifunctional photocatalytic structure where the strong, localized near field of the plasmonic metal nanoparticle (e.g., Ag) is coupled to the catalytic properties of the nonplasmonic metal nanoparticle (e.g., Pt) to enable chemical transformations. With an eye toward sustainable solar driven photocatalysis, we investigate how the structure of antenna-reactor complexes governs their photocatalytic activity in the light-limited regime, where all photons need to be effectively utilized. By synthesizing core@shell/satellite (Ag@SiO 2 /Pt) antenna-reactor complexes with varying Ag nanoparticle diameters and performing photocatalytic CO oxidation, we observed plasmon-enhanced photocatalysis only for antenna-reactor complexes with antenna components of intermediate sizes (25 and 50 nm). Optimal photocatalytic performance was shown to be determined by a balance between maximized local field enhancements at the catalytically active Pt surface, minimized collective scattering of photons out of the catalyst bed by the complexes, and minimal light absorption in the Ag nanoparticle antenna. These results elucidate the critical aspects of local field enhancement, light scattering, and absorption in plasmonic photocatalyst design, especially under light-limited illumination conditions.

Chemical Vapor Deposition of Turbine Thermal Barrier Coatings

NASA Technical Reports Server (NTRS)

Haven, Victor E.

1999-01-01

Ceramic thermal barrier coatings extend the operating temperature range of actively cooled gas turbine components, therefore increasing thermal efficiency. Performance and lifetime of existing ceram ic coatings are limited by spallation during heating and cooling cycles. Spallation of the ceramic is a function of its microstructure, which is determined by the deposition method. This research is investigating metalorganic chemical vapor deposition (MOCVD) of yttria stabilized zirconia to improve performance and reduce costs relative to electron beam physical vapor deposition. Coatings are deposited in an induction-heated, low-pressure reactor at 10 microns per hour. The coating's composition, structure, and response to the turbine environment will be characterized.

Total electron acceptor loading and composition affect hexavalent uranium reduction and microbial community structure in a membrane biofilm reactor.

PubMed

Ontiveros-Valencia, Aura; Zhou, Chen; Ilhan, Zehra Esra; de Saint Cyr, Louis Cornette; Krajmalnik-Brown, Rosa; Rittmann, Bruce E

2017-11-15

Molecular microbiology tools (i.e., 16S rDNA gene sequencing) were employed to elucidate changes in the microbial community structure according to the total electron acceptor loading (controlled by influent flow rate and/or medium composition) in a H 2 -based membrane biofilm reactor evaluated for removal of hexavalent uranium. Once nitrate, sulfate, and dissolved oxygen were replaced by U(VI) and bicarbonate and the total acceptor loading was lowered, slow-growing bacteria capable of reducing U(VI) to U(IV) dominated in the biofilm community: Replacing denitrifying bacteria Rhodocyclales and Burkholderiales were spore-producing Clostridiales and Natranaerobiales. Though potentially competing for electrons with U(VI) reducers, homo-acetogens helped attain steady U(VI) reduction, while methanogenesis inhibited U(VI) reduction. U(VI) reduction was reinstated through suppression of methanogenesis by addition of bromoethanesulfonate or by competition from SRB when sulfate was re-introduced. Predictive metagenome analysis further points out community changes in response to alterations in the electron-acceptor loading: Sporulation and homo-acetogenesis were critical factors for strengthening stable microbial U(VI) reduction. This study documents that sporulation was important to long-term U(VI) reduction, whether or not microorganisms that carry out U(VI) reduction mediated by cytochrome c 3 , such as SRB and ferric-iron-reducers, were inhibited. Copyright © 2017 Elsevier Ltd. All rights reserved.

Deleterious Thermal Effects Due To Randomized Flow Paths in Pebble Bed, and Particle Bed Style Reactors

NASA Technical Reports Server (NTRS)

Moran, Robert P.

2013-01-01

A review of literature associated with Pebble Bed and Particle Bed reactor core research has revealed a systemic problem inherent to reactor core concepts which utilize randomized rather than structured coolant channel flow paths. For both the Pebble Bed and Particle Bed Reactor designs; case studies reveal that for indeterminate reasons, regions within the core would suffer from excessive heating leading to thermal runaway and localized fuel melting. A thermal Computational Fluid Dynamics model was utilized to verify that In both the Pebble Bed and Particle Bed Reactor concepts randomized coolant channel pathways combined with localized high temperature regions would work together to resist the flow of coolant diverting it away from where it is needed the most to cooler less resistive pathways where it is needed the least. In other words given the choice via randomized coolant pathways the reactor coolant will take the path of least resistance, and hot zones offer the highest resistance. Having identified the relationship between randomized coolant channel pathways and localized fuel melting it is now safe to assume that other reactor concepts that utilize randomized coolant pathways such as the foam core reactor are also susceptible to this phenomenon.

Contributions from research on irradiated ferritic/martensitic steels to materials science and engineering

NASA Astrophysics Data System (ADS)

Gelles, D. S.

1990-05-01

Ferritic and martensitic steels are finding increased application for structural components in several reactor systems. Low-alloy steels have long been used for pressure vessels in light water fission reactors. Martensitic stainless steels are finding increasing usage in liquid metal fast breeder reactors and are being considered for fusion reactor applications when such systems become commercially viable. Recent efforts have evaluated the applicability of oxide dispersion-strengthened ferritic steels. Experiments on the effect of irradiation on these steels provide several examples where contributions are being made to materials science and engineering. Examples are given demonstrating improvements in basic understanding, small specimen test procedure development, and alloy development.

The prototype fast reactor at Dounreay, Scotland. Process and engineering development for sodium removal

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

Mann, A.; Herrick, R.; Gunn, J.

2007-07-01

Dounreay was home to commercial fast reactor development in the UK. Following the construction and operation of the Dounreay Fast Reactor, a sodium-cooled Prototype Fast Reactor (PFR), was constructed. PFR started operating in 1974, closed in 1994 and is presently being decommissioned. To date the bulk of the sodium has been removed and treated. Due to the design of the existing extraction system however, a sodium pool will remain in the heel of the reactor. To remove this sodium, a pump/camera system was developed, tested and deployed. The Water Vapour Nitrogen (WVN) process has been selected to allow removal ofmore » the final sodium residues from the reactor. Due to the design of the reactor and potential for structural damage should Normal WVN (which produces hydrated sodium hydroxide) be used, Low Concentration WVN (LC WVN) has been developed. Pilot scale testing has shown that it is possible treat the reactor within 18 months at a WVN concentration of up to 4% v/v and temperature of 120 deg. C. At present the equipment that will be used to apply LC WVN to the reactor is being developed at the detail design stage. and is expected to be deployed within the next few years. (authors)« less

Heat barrier for use in a nuclear reactor facility

DOEpatents

Keegan, Charles P.

1988-01-01

A thermal barrier for use in a nuclear reactor facility is disclosed herein. Generally, the thermal barrier comprises a flexible, heat-resistant web mounted over the annular space between the reactor vessel and the guard vessel in order to prevent convection currents generated in the nitrogen atmosphere in this space from entering the relatively cooler atmosphere of the reactor cavity which surrounds these vessels. Preferably, the flexible web includes a blanket of heat-insulating material formed from fibers of a refractory material, such as alumina and silica, sandwiched between a heat-resistant, metallic cloth made from stainless steel wire. In use, the web is mounted between the upper edges of the guard vessel and the flange of a sealing ring which surrounds the reactor vessel with a sufficient enough slack to avoid being pulled taut as a result of thermal differential expansion between the two vessels. The flexible web replaces the rigid and relatively complicated structures employed in the prior art for insulating the reactor cavity from the convection currents generated between the reactor vessel and the guard vessel.

Renewable Wood Pulp Paper Reactor with Hierarchical Micro/Nanopores for Continuous‐Flow Nanocatalysis

PubMed Central

Namba, Naoko; Takahashi, Tsukasa; Nogi, Masaya; Nishina, Yuta

2017-01-01

Abstract Continuous‐flow nanocatalysis based on metal nanoparticle catalyst‐anchored flow reactors has recently provided an excellent platform for effective chemical manufacturing. However, there has been limited progress in porous structure design and recycling systems for metal nanoparticle‐anchored flow reactors to create more efficient and sustainable catalytic processes. In this study, traditional paper is used for a highly efficient, recyclable, and even renewable flow reactor by tailoring the ultrastructures of wood pulp. The “paper reactor” offers hierarchically interconnected micro‐ and nanoscale pores, which can act as convective‐flow and rapid‐diffusion channels, respectively, for efficient access of reactants to metal nanoparticle catalysts. In continuous‐flow, aqueous, room‐temperature catalytic reduction of 4‐nitrophenol to 4‐aminophenol, a gold nanoparticle (AuNP)‐anchored paper reactor with hierarchical micro/nanopores provided higher reaction efficiency than state‐of‐the‐art AuNP‐anchored flow reactors. Inspired by traditional paper materials, successful recycling and renewal of AuNP‐anchored paper reactors were also demonstrated while high reaction efficiency was maintained. PMID:28394501

Quality assurance program requirements, Amendment 5 (9-26-79) to August 1973 issue

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

Not Available

This standard sets forth general requirements for planning, managing, conducting, and evaluating quality assurance programs for reactor development and test facility projects and associated processes, structures, components, and systems. These quality assurance requirements are based on proven practices and provide the means of control and verification whereby those responsible fo poject management can assure that the quality required for safe, reliable, and economical operation will be achieved. The objective of the program of the programs covered by this standard is to assure that structures, components, systems, and facilities are designed, developed, manufactured, constructed, operated, and maintained in compliance with establishedmore » engineering criteria. To achieve this objective, controls are to be established and implemented at predetermined points, and necessary action taken to prevent, detect, and correct any deficiencies.« less

A simulator-based nuclear reactor emergency response training exercise.

PubMed

Waller, Edward; Bereznai, George; Shaw, John; Chaput, Joseph; Lafortune, Jean-Francois

Training offsite emergency response personnel basic awareness of onsite control room operations during nuclear power plant emergency conditions was the primary objective of a week-long workshop conducted on a CANDU® virtual nuclear reactor simulator available at the University of Ontario Institute of Technology, Oshawa, Canada. The workshop was designed to examine both normal and abnormal reactor operating conditions, and to observe the conditions in the control room that may have impact on the subsequent offsite emergency response. The workshop was attended by participants from a number of countries encompassing diverse job functions related to nuclear emergency response. Objectives of the workshop were to provide opportunities for participants to act in the roles of control room personnel under different reactor operating scenarios, providing a unique experience for participants to interact with the simulator in real-time, and providing increased awareness of control room operations during accident conditions. The ability to "pause" the simulator during exercises allowed the instructors to evaluate and critique the performance of participants, and to provide context with respect to potential offsite emergency actions. Feedback from the participants highlighted (i) advantages of observing and participating "hands-on" with operational exercises, (ii) their general unfamiliarity with control room operational procedures and arrangements prior to the workshop, (iii) awareness of the vast quantity of detailed control room procedures for both normal and transient conditions, and (iv) appreciation of the increased workload for the operators in the control room during a transient from normal operations. Based upon participant feedback, it was determined that the objectives of the training had been met, and that future workshops should be conducted.

Effect Of Organic Substrate Composition On Microbial Community Structure Of Pilot-Scale Biochemical Reactors Treating Mining Influenced Water

EPA Science Inventory

Mining-influenced water (MIW) is acidic, metal rich water formed when sulfide minerals react with oxygen and water. There are various options for the treatment of MIW; however, passive biological systems such as biochemical reactors (BCRs) have shown promise because of their low...

Effect Of Organic Substrate Composition On Microbial Community Structure Of Pilot-Scale Biochemical Reactors Treating Mining Influenced Water - (Presentation)

EPA Science Inventory

Mining-influenced water (MIW) is acidic, metal rich water formed when sulfide minerals react with oxygen and water. There are various options for the treatment of MIW; however, passive biological systems such as biochemical reactors (BCRs) have shown promise because of their low...

Reactors Save Energy, Costs for Hydrogen Production

NASA Technical Reports Server (NTRS)

2014-01-01

While examining fuel-reforming technology for fuel cells onboard aircraft, Glenn Research Center partnered with Garrettsville, Ohio-based Catacel Corporation through the Glenn Alliance Technology Exchange program and a Space Act Agreement. Catacel developed a stackable structural reactor that is now employed for commercial hydrogen production and results in energy savings of about 20 percent.

Selective synthesis of human milk fat-style structured triglycerides from microalgal oil in a microfluidic reactor packed with immobilized lipase

DOE PAGES

Wang, Jun; Liu, Xi; Wang, Xu -Dong; ...

2016-08-18

Human milk fat-style structured triacylglycerols were produced from microalgal oil in a continuous microfluidic reactor packed with immobilized lipase for the first time. A remarkably high conversion efficiency was demonstrated in the microreactor with reaction time being reduced by 8 times, Michaelis constant decreased 10 times, the lipase reuse times increased 2.25-fold compared to those in a batch reactor. In addition, the content of palmitic acid at sn-2 position (89.0%) and polyunsaturated fatty acids at sn-1, 3 positions (81.3%) are slightly improved compared to the product in a batch reactor. The increase of melting points (1.7 °C) and decrease ofmore » crystallizing point (3 °C) implied higher quality product was produced using the microfluidic technology. The main cost can be reduced from 212.3 to 14.6 per batch with the microreactor. Altogether, the microfluidic bioconversion technology is promising for modified functional lipids production allowing for cost-effective approach to produce high-value microalgal coproducts.« less

Selective synthesis of human milk fat-style structured triglycerides from microalgal oil in a microfluidic reactor packed with immobilized lipase

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

Wang, Jun; Liu, Xi; Wang, Xu -Dong

Human milk fat-style structured triacylglycerols were produced from microalgal oil in a continuous microfluidic reactor packed with immobilized lipase for the first time. A remarkably high conversion efficiency was demonstrated in the microreactor with reaction time being reduced by 8 times, Michaelis constant decreased 10 times, the lipase reuse times increased 2.25-fold compared to those in a batch reactor. In addition, the content of palmitic acid at sn-2 position (89.0%) and polyunsaturated fatty acids at sn-1, 3 positions (81.3%) are slightly improved compared to the product in a batch reactor. The increase of melting points (1.7 °C) and decrease ofmore » crystallizing point (3 °C) implied higher quality product was produced using the microfluidic technology. The main cost can be reduced from 212.3 to 14.6 per batch with the microreactor. Altogether, the microfluidic bioconversion technology is promising for modified functional lipids production allowing for cost-effective approach to produce high-value microalgal coproducts.« less

Hyperthermal Environments Simulator for Nuclear Rocket Engine Development

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Foote, John P.; Clifton, W. B.; Hickman, Robert R.; Wang, Ten-See; Dobson, Christopher C.

2011-01-01

An arc-heater driven hyperthermal convective environments simulator was recently developed and commissioned for long duration hot hydrogen exposure of nuclear thermal rocket materials. This newly established non-nuclear testing capability uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce hightemperature pressurized hydrogen flows representative of nuclear reactor core environments, excepting radiation effects, and is intended to serve as a low-cost facility for supporting non-nuclear developmental testing of hightemperature fissile fuels and structural materials. The resulting reactor environments simulator represents a valuable addition to the available inventory of non-nuclear test facilities and is uniquely capable of investigating and characterizing candidate fuel/structural materials, improving associated processing/fabrication techniques, and simulating reactor thermal hydraulics. This paper summarizes facility design and engineering development efforts and reports baseline operational characteristics as determined from a series of performance mapping and long duration capability demonstration tests. Potential follow-on developmental strategies are also suggested in view of the technical and policy challenges ahead. Keywords: Nuclear Rocket Engine, Reactor Environments, Non-Nuclear Testing, Fissile Fuel Development.

Cleanup Verification Package for the 118-H-6:2, 105-H Reactor Ancillary Support Areas, Below-Grade Structures, and Underlying Soils; the 118-H-6:3, 105-H Reactor Fuel Storage Basin and Underlying Soils; The 118-H-6:3 Fuel Storage Basin Deep Zone Side Slope Soils; the 100-H-9, 100-H-10, and 100-H-13 French Drains; the 100-H-11 and 100-H-12 Expansion Box French Drains; and the 100-H-14 and 100-H-31 Surface Contamination Zones

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

M. J. Appel

2006-06-29

This cleanup verification package documents completion of removal actions for the 105-H Reactor Ancillary Support Areas, Below-Grade Structures, and Underlying Soils (subsite 118-H-6:2); 105-H Reactor Fuel Storage Basin and Underlying Soils (118-H-6:3); and Fuel Storage Basin Deep Zone Side Slope Soils. This CVP also documents remedial actions for the following seven additional waste sties: French Drain C (100-H-9), French Drain D (100-H-10), Expansion Box French Drain E (100-H-11), Expansion Box French Drain F (100-H-12), French Drain G (100-H-13), Surface Contamination Zone H (100-H-14), and the Polychlorinated Biphenyl Surface Contamination Zone (100-H-31).

Selective synthesis of human milk fat-style structured triglycerides from microalgal oil in a microfluidic reactor packed with immobilized lipase.

PubMed

Wang, Jun; Liu, Xi; Wang, Xu-Dong; Dong, Tao; Zhao, Xing-Yu; Zhu, Dan; Mei, Yi-Yuan; Wu, Guo-Hua

2016-11-01

Human milk fat-style structured triacylglycerols were produced from microalgal oil in a continuous microfluidic reactor packed with immobilized lipase for the first time. A remarkably high conversion efficiency was demonstrated in the microreactor with reaction time being reduced by 8 times, Michaelis constant decreased 10 times, the lipase reuse times increased 2.25-fold compared to those in a batch reactor. In addition, the content of palmitic acid at sn-2 position (89.0%) and polyunsaturated fatty acids at sn-1, 3 positions (81.3%) are slightly improved compared to the product in a batch reactor. The increase of melting points (1.7°C) and decrease of crystallizing point (3°C) implied higher quality product was produced using the microfluidic technology. The main cost can be reduced from $212.3 to $14.6 per batch with the microreactor. Overall, the microfluidic bioconversion technology is promising for modified functional lipids production allowing for cost-effective approach to produce high-value microalgal coproducts. Copyright © 2016 Elsevier Ltd. All rights reserved.

Performance enhanced piezoelectric-based crack detection system for high temperature I-beam SHM

NASA Astrophysics Data System (ADS)

Zhang, Chen; Zhang, Haifeng

2017-04-01

This paper proposes an innovative sensing system for high temperature (up to 150°C) I-beam crack detection. The proposed system is based on the piezoelectric effect and laser sensing mechanisms, which is proved to be effective at high temperature environment (up to 150°C). Different from other high temperature SHM approaches, the proposed sensing system is employing a piezoelectric disk as an actuator and a laser vibrometer as a sensor for remote detection. Lab tests are carried out and the vibrational properties are studied to characterize the relationship between crack depth and sensor responses by analyzing the RMS of sensor responses. Instead of utilizing a pair of piezoelectric actuator and sensor, using the laser vibrometer will enable 1) a more flexible detection, which will not be limited to specific area or dimension, 2) wireless sensing, which lowers the risk of operating at high temperature/harsh environment. The proposed sensing system can be applied to engineering structures such as in nuclear power plant reactor vessel and heat pipe structures/systems.

Designing Reactor Microbiomes for Chemical Production from Organic Waste.

PubMed

Oleskowicz-Popiel, Piotr

2018-01-27

Microorganisms are responsible for biochemical cycles and therefore play essential roles in the environment. By using omics approaches and network analysis to understand the interaction and cooperation within mixed microbial communities, it would be possible to engineer microbiomes in fermentation and digestion reactors to convert organic waste into valuable products. Copyright © 2018 Elsevier Ltd. All rights reserved.

Problems and Delays Overshadow NRC's Initial Success in Improving Reactor Operators' Capabilities.

ERIC Educational Resources Information Center

General Accounting Office, Washington, DC.

The nuclear power plant accident at Three Mile Island raised many questions concerning the safety of nuclear power plant operations and the ability of nuclear plant reactor operators to respond to abnormal or accident conditions. In response, the Nuclear Regulatory Commission (NRC) developed a plan, which included short- and long-term actions to…

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.

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.

Radiobiologic significance of response of intratumor quiescent cells in vivo to accelerated carbon ion beams compared with gamma-rays and reactor neutron beams.

PubMed

Masunaga, Shin-Ichiro; Ando, Koichi; Uzawa, Akiko; Hirayama, Ryoichi; Furusawa, Yoshiya; Koike, Sachiko; Sakurai, Yoshinori; Nagata, Kenji; Suzuki, Minoru; Kashino, Genro; Kinashi, Yuko; Tanaka, Hiroki; Maruhashi, Akira; Ono, Koji

2008-01-01

To clarify the radiosensitivity of intratumor quiescent cells in vivo to accelerated carbon ion beams and reactor neutron beams. Squamous cell carcinoma VII tumor-bearing mice were continuously given 5-bromo-2'-deoxyuridine to label all intratumor proliferating cells. Next, they received accelerated carbon ion or gamma-ray high-dose-rate (HDR) or reduced-dose-rate (RDR) irradiation. Other tumor-bearing mice received reactor thermal or epithermal neutrons with RDR irradiation. Immediately after HDR and RDR irradiation or 12 h after HDR irradiation, the response of quiescent cells was assessed in terms of the micronucleus frequency using immunofluorescence staining for 5-bromo-2'-deoxyuridine. The response of the total (proliferating plus quiescent) tumor cells was determined from the 5-bromo-2'-deoxyuridine nontreated tumors. The difference in radiosensitivity between the total and quiescent cell populations after gamma-ray irradiation was markedly reduced with reactor neutron beams or accelerated carbon ion beams, especially with a greater linear energy transfer (LET) value. Clearer repair in quiescent cells than in total cells through delayed assay or a decrease in the dose rate with gamma-ray irradiation was efficiently inhibited with carbon ion beams, especially with a greater LET. With RDR irradiation, the radiosensitivity to accelerated carbon ion beams with a greater LET was almost similar to that to reactor thermal and epithermal neutron beams. In terms of tumor cell-killing effect as a whole, including quiescent cells, accelerated carbon ion beams, especially with greater LET values, are very useful for suppressing the dependency on the heterogeneity within solid tumors, as well as depositing the radiation dose precisely.

Development of an advanced antineutrino detector for reactor monitoring

DOE PAGES

Classen, T.; Bernstein, A.; Bowden, N. S.; ...

2014-11-05

We present the development of a compact antineutrino detector for the purpose of nuclear reactor monitoring, improving upon a previously successful design. Our paper will describe the design improvements of the detector which increases the antineutrino detection efficiency threefold over the previous effort. There are two main design improvements over previous generations of detectors for nuclear reactor monitoring: dual-ended optical readout and single volume detection mass. The dual-ended optical readout eliminates the need for fiducialization and increases the uniformity of the detector's optical response. The containment of the detection mass in a single active volume provides more target mass permore » detector footprint, a key design criteria for operating within a nuclear power plant. This technology could allow for real-time monitoring of the evolution of a nuclear reactor core, independent of reactor operator declarations of fuel inventories, and may be of interest to the safeguards community.« less

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

Pind, C.

The SECURE heating reactor was designed by ASEA-ATOM as a realistic alternative for district heating in urban areas and for supplying heat to process industries. SECURE has unique safety characteristics, that are based on fundamental laws of physics. The safety does not depend on active components or operator intervention for shutdown and cooling of the reactor. The inherent safety characteristics of the plant cannot be affected by operator errors. Due to its very low environment impact, it can be sited close to heat consumers. The SECURE heating reactor has been shown to be competitive in comparison with other alternatives formore » heating Helsinki and Seoul. The SECURE heating reactor forms a basis for the power-producing SECURE-P reactor known as PIUS (Process Inherent Ultimate Safety), which is based on the same inherent safety principles. The thermohydraulic function and transient response have been demonstrated in a large electrically heated loop at the ASEA-ATOM laboratories.« less

Study of the diversity of microbial communities in a sequencing batch reactor oxic-settling-anaerobic process and its modified process.

PubMed

Sun, Lianpeng; Chen, Jianfan; Wei, Xiange; Guo, Wuzhen; Lin, Meishan; Yu, Xiaoyu

2016-05-01

To further reveal the mechanism of sludge reduction in the oxic-settling-anaerobic (OSA) process, the polymerase chain reaction - denaturing gradient gel electrophoresis protocol was used to study the possible difference in the microbial communities between a sequencing batch reactor (SBR)-OSA process and its modified process, by analyzing the change in the diversity of the microbial communities in each reactor of both systems. The results indicated that the structure of the microbial communities in aerobic reactors of the 2 processes was very different, but the predominant microbial populations in anaerobic reactors were similar. The predominant microbial population in the aerobic reactor of the SBR-OSA belonged to Burkholderia cepacia, class Betaproteobacteria, while those of the modified process belonged to the classes Alphaproteobacteria, Betaproteobacteria, and Gammaproteobacteria. These 3 types of microbes had a cryptic growth characteristic, which was the main cause of a greater sludge reduction efficiency achieved by the modified process.

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.

Superfund record of decision (EPA region 10): Idaho National Engineering Lab, (USDOE) Operable Unit 26 (Stationary Low-Power Reactor-1 and Boiling Water Reactor Experiment-I Burial Grounds), Idaho Falls, ID, December 1, 1995

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

NONE

1997-03-01

This document presents the selected remedial action for the Stationary Low-Power Reactor-1 (SL-1) burial ground, the Boiling Water Reactor Experiment-I (BORAX-I) burial ground, and 10 no action sites in Waste Area Group 5. Actual or threatened releases of hazardous substances from the SL-1 and BORAX-I burial grounds, if not addressed by implementing the response action selected in this Record of Decision, may present a current or potential threat to public health, welfare, or the environment. The 10 no action sites do not present a threat to human health or the environment.

An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

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

Boing, L.E.; Henley, D.R.; Manion, W.J.

1989-12-01

Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document inmore » their evaluation process. 73 refs., 26 figs., 69 tabs.« less

Corn stalks char from fast pyrolysis as precursor material for preparation of activated carbon in fluidized bed reactor.

PubMed

Wang, Zhiqi; Wu, Jingli; He, Tao; Wu, Jinhu

2014-09-01

Corn stalks char from fast pyrolysis was activated by physical and chemical activation process in a fluidized bed reactor. The structure and morphology of the carbons were characterized by N2 adsorption and SEM. Effects of activation time and activation agents on the structure of activation carbon were investigated. The physically activated carbons with CO2 have BET specific surface area up to 880 m(2)/g, and exhibit microporous structure. The chemically activated carbons with H3PO4 have BET specific surface area up to 600 m(2)/g, and exhibit mesoporous structure. The surface morphology shows that physically activated carbons exhibit fibrous like structure in nature with long ridges, resembling parallel lines. Whereas chemically activated carbons have cross-interconnected smooth open pores without the fibrous like structure. Copyright © 2014 Elsevier Ltd. All rights reserved.

Nuclear transmutation in steels

NASA Astrophysics Data System (ADS)

Belozerova, A. R.; Shimanskii, G. A.; Belozerov, S. V.

2009-05-01

The investigations of the effects of nuclear transmutation in steels that are widely used in nuclear power and research reactors and in steels that are planned for the application in thermonuclear fusion plants, which are employed under the conditions of a prolonged action of neutron irradiation with different spectra, made it possible to study the effects of changes in the isotopic and chemical composition on the tendency of changes in the structural stability of these steels. For the computations of nuclear transmutation in steels, we used a program complex we have previously developed on the basis of algorithms for constructing branched block-type diagrams of nuclide transformations and for locally and globally optimizing these diagrams with the purpose of minimizing systematic errors in the calculation of nuclear transmutation. The dependences obtained were applied onto a Schaeffler diagram for steels used for structural elements of reactors. For the irradiation in fission reactors, we observed only a weak influence of the effects of nuclear transmutation in steels on their structural stability. On the contrary, in the case of irradiation with fusion neutrons, a strong influence of the effects of nuclear transmutation in steels on their structural stability has been noted.

Solid state oxygen anion and electron mediating membrane and catalytic membrane reactors containing them

DOEpatents

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

2005-09-27

This invention relates to gas-impermeable, solid state materials fabricated into membranes for use in catalytic membrane reactors. This invention particularly relates to solid state oxygen anion- and electron-mediating membranes for use in catalytic membrane reactors for promoting partial or full oxidation of different chemical species, for decomposition of oxygen-containing species, and for separation of oxygen from other gases. Solid state materials for use in the membranes of this invention include mixed metal oxide compounds having the brownmillerite crystal structure.

Solid state oxygen anion and electron mediating membrane and catalytic membrane reactors containing them

DOEpatents

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

2000-01-01

This invention relates to gas-impermeable, solid state materials fabricated into membranes for use in catalytic membrane reactors. This invention particularly relates to solid state oxygen anion- and electron-mediating membranes for use in catalytic membrane reactors for promoting partial or full oxidation of different chemical species, for decomposition of oxygen-containing species, and for separation of oxygen from other gases. Solid state materials for use in the membranes of this invention include mixed metal oxide compounds having the brownmillerite crystal structure.

Application of a fluidized bed reactor charged with aragonite for control of alkalinity, pH and carbon dioxide in marine recirculating aquaculture systems

USGS Publications Warehouse

Paul S Wills, PhD; Pfeiffer, Timothy; Baptiste, Richard; Watten, Barnaby J.

2016-01-01

Control of alkalinity, dissolved carbon dioxide (dCO2), and pH are critical in marine recirculating aquaculture systems (RAS) in order to maintain health and maximize growth. A small-scale prototype aragonite sand filled fluidized bed reactor was tested under varying conditions of alkalinity and dCO2 to develop and model the response of dCO2 across the reactor. A large-scale reactor was then incorporated into an operating marine recirculating aquaculture system to observe the reactor as the system moved toward equilibrium. The relationship between alkalinity dCO2, and pH across the reactor are described by multiple regression equations. The change in dCO2 across the small-scale reactor indicated a strong likelihood that an equilibrium alkalinity would be maintained by using a fluidized bed aragonite reactor. The large-scale reactor verified this observation and established equilibrium at an alkalinity of approximately 135 mg/L as CaCO3, dCO2 of 9 mg/L, and a pH of 7.0 within 4 days that was stable during a 14 day test period. The fluidized bed aragonite reactor has the potential to simplify alkalinity and pH control, and aid in dCO2 control in RAS design and operation. Aragonite sand, purchased in bulk, is less expensive than sodium bicarbonate and could reduce overall operating production costs.

Evaluation of ilmenite serpentine concrete and ordinary concrete as nuclear reactor shielding

NASA Astrophysics Data System (ADS)

Abulfaraj, Waleed H.; Kamal, Salah M.

1994-07-01

The present study involves adapting a formal decision methodology to the selection of alternative nuclear reactor concretes shielding. Multiattribute utility theory is selected to accommodate decision makers' preferences. Multiattribute utility theory (MAU) is here employed to evaluate two appropriate nuclear reactor shielding concretes in terms of effectiveness to determine the optimal choice in order to meet the radiation protection regulations. These concretes are Ordinary concrete (O.C.) and Ilmenite Serpentile concrete (I.S.C.). These are normal weight concrete and heavy heat resistive concrete, respectively. The effectiveness objective of the nuclear reactor shielding is defined and structured into definite attributes and subattributes to evaluate the best alternative. Factors affecting the decision are dose received by reactor's workers, the material properties as well as cost of concrete shield. A computer program is employed to assist in performing utility analysis. Based upon data, the result shows the superiority of Ordinary concrete over Ilmenite Serpentine concrete.

High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation

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

Was, Gary; Wirth, Brian; Motta, Athur

The objective of this proposal is to demonstrate the capability to predict the evolution of microstructure and properties of structural materials in-reactor and at high doses, using ion irradiation as a surrogate for reactor irradiations. “Properties” includes both physical properties (irradiated microstructure) and the mechanical properties of the material. Demonstration of the capability to predict properties has two components. One is ion irradiation of a set of alloys to yield an irradiated microstructure and corresponding mechanical behavior that are substantially the same as results from neutron exposure in the appropriate reactor environment. Second is the capability to predict the irradiatedmore » microstructure and corresponding mechanical behavior on the basis of improved models, validated against both ion and reactor irradiations and verified against ion irradiations. Taken together, achievement of these objectives will yield an enhanced capability for simulating the behavior of materials in reactor irradiations.« less

GAS COOLED NUCLEAR REACTORS

DOEpatents

Long, E.; Rodwell, W.

1958-06-10

A gas-cooled nuclear reactor consisting of a graphite reacting core and reflector structure supported in a containing vessel is described. A gas sealing means is included for sealing between the walls of the graphite structure and containing vessel to prevent the gas coolant by-passing the reacting core. The reacting core is a multi-sided right prismatic structure having a pair of parallel slots around its periphery. The containing vessel is cylindrical and has a rib on its internal surface which supports two continuous ring shaped flexible web members with their radially innermost ends in sealing engagement within the radially outermost portion of the slots. The core structure is supported on ball bearings. This design permits thermal expansion of the core stracture and vessel while maintainirg a peripheral seal between the tvo elements.

Standard interface files and procedures for reactor physics codes, version III

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

Carmichael, B.M.

Standards and procedures for promoting the exchange of reactor physics codes are updated to Version-III status. Standards covering program structure, interface files, file handling subroutines, and card input format are included. The implementation status of the standards in codes and the extension of the standards to new code areas are summarized. (15 references) (auth)

HEAT EXCHANGER BUILDING, TRA644. NORTHEAST CORNER. CAMERA IS ON PIKE ...

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

HEAT EXCHANGER BUILDING, TRA-644. NORTHEAST CORNER. CAMERA IS ON PIKE STREET FACING SOUTHWEST. ATTACHED STRUCTURE AT RIGHT OF VIEW IS ETR COMPRESSOR BUILDING, TRA-643. INL NEGATIVE NO. HD46-36-4. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

Leskovar, Matjaz; Koncar, Bostjan

An ex-vessel steam explosion may occur when during a severe reactor accident the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles at later times, during the expansion of the highly pressurized water vapor, that may endanger surrounding structures. In contrast to specialized steammore » explosion CFD codes, where the steam explosion is modeled on micro-scale using fundamental averaged multiphase flow conservation equations, in the presented approach the steam explosion is modeled in a simplified manner as an expanding high-pressure pre-mixture of dispersed molten fuel, liquid water and vapor. Applying the developed steam explosion model, a comprehensive analysis of the ex-vessel steam explosion in a typical PWR reactor cavity was done using the CFD code CFX-10. At four selected locations, which are of importance for the assessment of the vulnerability of cavity structures, the pressure histories were recorded and the corresponding pressure impulses calculated. The pressure impulses determine the destructive potential of the steam explosion and represent the input for the structural mechanical analysis of the cavity structures. The simulation results show that the pressure impulses depend mainly on the steam explosion energy conversion ratio, whereas the influence of the pre-mixture vapor volume fraction, which is a parameter in our model and determines the maximum steam explosion pressure, is not significant. (authors)« less

Systems of mechanized and reactive droplets powered by multi-responsive surfactants

NASA Astrophysics Data System (ADS)

Yang, Zhijie; Wei, Jingjing; Sobolev, Yaroslav I.; Grzybowski, Bartosz A.

2018-01-01

Although ‘active’ surfactants, which are responsive to individual external stimuli such as temperature, electric or magnetic fields, light, redox processes or chemical agents, are well known, it would be interesting to combine several of these properties within one surfactant species. Such multi-responsive surfactants could provide ways of manipulating individual droplets and possibly assembling them into larger systems of dynamic reactors. Here we describe surfactants based on functionalized nanoparticle dimers that combine all of these and several other characteristics. These surfactants and therefore the droplets that they cover are simultaneously addressable by magnetic, optical and electric fields. As a result, the surfactant-covered droplets can be assembled into various hierarchical structures, including dynamic ones, in which light powers the rapid rotation of the droplets. Such rotating droplets can transfer mechanical torques to their non-nearest neighbours, thus acting like systems of mechanical gears. Furthermore, droplets of different types can be merged by applying electric fields and, owing to interfacial jamming, can form complex, non-spherical, ‘patchy’ structures with different surface regions covered with different surfactants. In systems of droplets that carry different chemicals, combinations of multiple stimuli can be used to control the orientations of the droplets, inter-droplet transport, mixing of contents and, ultimately, sequences of chemical reactions. Overall, the multi-responsive active surfactants that we describe provide an unprecedented level of flexibility with which liquid droplets can be manipulated, assembled and reacted.

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

Li, C.; Yu, G.; Wang, K.

The physical designs of the new concept reactors which have complex structure, various materials and neutronic energy spectrum, have greatly improved the requirements to the calculation methods and the corresponding computing hardware. Along with the widely used parallel algorithm, heterogeneous platforms architecture has been introduced into numerical computations in reactor physics. Because of the natural parallel characteristics, the CPU-FPGA architecture is often used to accelerate numerical computation. This paper studies the application and features of this kind of heterogeneous platforms used in numerical calculation of reactor physics through practical examples. After the designed neutron diffusion module based on CPU-FPGA architecturemore » achieves a 11.2 speed up factor, it is proved to be feasible to apply this kind of heterogeneous platform into reactor physics. (authors)« less

The effects of electric power industry restructuring on the safety of nuclear power plants in the United States

NASA Astrophysics Data System (ADS)

Butler, Thomas S.

Throughout the United States the electric utility industry is restructuring in response to federal legislation mandating deregulation. The electric utility industry has embarked upon an extraordinary experiment by restructuring in response to deregulation that has been advocated on the premise of improving economic efficiency by encouraging competition in as many sectors of the industry as possible. However, unlike the telephone, trucking, and airline industries, the potential effects of electric deregulation reach far beyond simple energy economics. This dissertation presents the potential safety risks involved with the deregulation of the electric power industry in the United States and abroad. The pressures of a competitive environment on utilities with nuclear power plants in their portfolio to lower operation and maintenance costs could squeeze them to resort to some risky cost-cutting measures. These include deferring maintenance, reducing training, downsizing staff, excessive reductions in refueling down time, and increasing the use of on-line maintenance. The results of this study indicate statistically significant differences at the .01 level between the safety of pressurized water reactor nuclear power plants and boiling water reactor nuclear power plants. Boiling water reactors exhibited significantly more problems than did pressurized water reactors.

New reactor cavity cooling system having passive safety features using novel shape for HTGRs and VHTRs

DOE PAGES

Takamatsu, Kuniyoshi; Hu, Rui

2014-11-27

A new, highly efficient reactor cavity cooling system (RCCS) with passive safety features without a requirement for electricity and mechanical drive is proposed for high temperature gas cooled reactors (HTGRs) and very high temperature reactors (VHTRs). The RCCS design consists of continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40 (°C). The RCCS uses a novel shape to efficiently remove the heat released from the RPV with radiation and natural convection. Employing the air as the working fluid and the ambient airmore » as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. Therefore, HTGRs and VHTRs adopting the new RCCS design can avoid core melting due to overheating the fuels. The simulation results from a commercial CFD code, STAR-CCM+, show that the temperature distribution of the RCCS is within the temperature limits of the structures, such as the maximum operating temperature of the RPV, 713.15 (K) = 440 (°C), and the heat released from the RPV could be removed safely, even during a loss of coolant accident (LOCA). Finally, when the RCCS can remove 600 (kW) of the rated nominal state even during LOCA, the safety review for building the HTTR could confirm that the temperature distribution of the HTTR is within the temperature limits of the structures to secure structures and fuels after the shutdown because the large heat capacity of the graphite core can absorb heat from the fuel in a short period. Therefore, the capacity of the new RCCS design would be sufficient for decay heat removal.« less

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

Natesan, K.; Momozaki, Y.; Li, M.

This report gives a description of the activities in design, fabrication, construction, and assembling of a pumped sodium loop for the sodium compatibility studies on advanced structural materials. The work is the Argonne National Laboratory (ANL) portion of the effort on the work project entitled, 'Sodium Compatibility of Advanced Fast Reactor Materials,' and is a part of Advanced Materials Development within the Reactor Campaign. The objective of this project is to develop information on sodium corrosion compatibility of advanced materials being considered for sodium reactor applications. This report gives the status of the sodium pumped loop at Argonne National Laboratory,more » the specimen details, and the technical approach to evaluate the sodium compatibility of advanced structural alloys. This report is a deliverable from ANL in FY2010 (M2GAN10SF050302) under the work package G-AN10SF0503 'Sodium Compatibility of Advanced Fast Reactor Materials.' Two reports were issued in 2009 (Natesan and Meimei Li 2009, Natesan et al. 2009) which examined the thermodynamic and kinetic factors involved in the purity of liquid sodium coolant for sodium reactor applications as well as the design specifications for the ANL pumped loop for testing advanced structural materials. Available information was presented on solubility of several metallic and nonmetallic elements along with a discussion of the possible mechanisms for the accumulation of impurities in sodium. That report concluded that the solubility of many metals in sodium is low (<1 part per million) in the temperature range of interest in sodium reactors and such trace amounts would not impact the mechanical integrity of structural materials and components. The earlier report also analyzed the solubility and transport mechanisms of nonmetallic elements such as oxygen, nitrogen, carbon, and hydrogen in laboratory sodium loops and in reactor systems such as Experimental Breeder Reactor-II, Fast Flux Test Facility, and Clinch River Breeder Reactor. Among the nonmetallic elements discussed, oxygen is deemed controllable and its concentration in sodium can be maintained in sodium for long reactor life by using cold-trap method. It was concluded that among the cold-trap and getter-trap methods, the use of cold trap is sufficient to achieve oxygen concentration of the order of 1 part per million. Under these oxygen conditions in sodium, the corrosion performance of structural materials such as austenitic stainless steels and ferritic steels will be acceptable at a maximum core outlet sodium temperature of {approx}550 C. In the current sodium compatibility studies, the oxygen concentration in sodium will be controlled and maintained at {approx}1 ppm by controlling the cold trap temperature. The oxygen concentration in sodium in the forced convection sodium loop will be controlled and monitored by maintaining the cold trap temperature in the range of 120-150 C, which would result in oxygen concentration in the range of 1-2 ppm. Uniaxial tensile specimens are being exposed to flowing sodium and will be retrieved and analyzed for corrosion and post-exposure tensile properties. Advanced materials for sodium exposure include austenitic alloy HT-UPS and ferritic-martensitic steels modified 9Cr-1Mo and NF616. Among the nonmetallic elements in sodium, carbon was assessed to have the most influence on structural materials since carbon, as an impurity, is not amenable to control and maintenance by any of the simple purification methods. The dynamic equilibrium value for carbon in sodium systems is dependent on several factors, details of which were discussed in the earlier report. The current sodium compatibility studies will examine the role of carbon concentration in sodium on the carburization-decarburization of advanced structural materials at temperatures up to 650 C. Carbon will be added to the sodium by exposure of carbon-filled iron tubes, which over time will enable carbon to diffuse through iron and dissolve into sodium. The method enables addition of dissolved carbon (without carbon particulates) in sodium that is of interest for materials compatibility evaluation. The removal of carbon from the sodium will be accomplished by exposing carbon-gettering alloys such as refractory metals that have a high partitioning coefficient for carbon and also precipitate carbides, thereby decreasing the carbon concentration in sodium.« less

Calibration of ITER Instant Power Neutron Monitors: Recommended Scenario of Experiments at the Reactor

NASA Astrophysics Data System (ADS)

Borisov, A. A.; Deryabina, N. A.; Markovskij, D. V.

2017-12-01

Instant power is a key parameter of the ITER. Its monitoring with an accuracy of a few percent is an urgent and challenging aspect of neutron diagnostics. In a series of works published in Problems of Atomic Science and Technology, Series: Thermonuclear Fusion under a common title, the step-by-step neutronics analysis was given to substantiate a calibration technique for the DT and DD modes of the ITER. A Gauss quadrature scheme, optimal for processing "expensive" experiments, is used for numerical integration of 235U and 238U detector responses to the point sources of 14-MeV neutrons. This approach allows controlling the integration accuracy in relation to the number of coordinate mesh points and thus minimizing the number of irradiations at the given uncertainty of the full monitor response. In the previous works, responses of the divertor and blanket monitors to the isotropic point sources of DT and DD neutrons in the plasma profile and to the models of real sources were calculated within the ITER model using the MCNP code. The neutronics analyses have allowed formulating the basic principles of calibration that are optimal for having the maximum accuracy at the minimum duration of in situ experiments at the reactor. In this work, scenarios of the preliminary and basic experimental ITER runs are suggested on the basis of those principles. It is proposed to calibrate the monitors only with DT neutrons and use correction factors to the DT mode calibration for the DD mode. It is reasonable to perform full calibration only with 235U chambers and calibrate 238U chambers by responses of the 235U chambers during reactor operation (cross-calibration). The divertor monitor can be calibrated using both direct measurement of responses at the Gauss positions of a point source and simplified techniques based on the concepts of equivalent ring sources and inverse response distributions, which will considerably reduce the amount of measurements. It is shown that the monitor based on the average responses of the horizontal and vertical neutron chambers remains spatially stable as the source moves and can be used in addition to the staff monitor at neutron fluxes in the detectors four orders of magnitude lower than on the first wall, where staff detectors are located. Owing to low background, detectors of neutron chambers do not need calibration in the reactor because it is actually determination of the absolute detector efficiency for 14-MeV neutrons, which is a routine out-of-reactor procedure.

Dynamic analysis of gas-core reactor system

NASA Technical Reports Server (NTRS)

Turner, K. H., Jr.

1973-01-01

A heat transfer analysis was incorporated into a previously developed model CODYN to obtain a model of open-cycle gaseous core reactor dynamics which can predict the heat flux at the cavity wall. The resulting model was used to study the sensitivity of the model to the value of the reactivity coefficients and to determine the system response for twenty specified perturbations. In addition, the model was used to study the effectiveness of several control systems in controlling the reactor. It was concluded that control drums located in the moderator region capable of inserting reactivity quickly provided the best control.

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.

Anaerobic digestion for simultaneous sewage sludge treatment and CO biomethanation: process performance and microbial ecology.

PubMed

Luo, Gang; Wang, Wen; Angelidaki, Irini

2013-09-17

Syngas is produced by thermal gasification of both nonrenewable and renewable sources including biomass and coal, and it consists mainly of CO, CO2, and H2. In this paper we aim to bioconvert CO in the syngas to CH4. A novel technology for simultaneous sewage sludge treatment and CO biomethanation in an anaerobic reactor was presented. Batch experiments showed that CO was inhibitory to methanogens, but not to bacteria, at CO partial pressure between 0.25 and 1 atm under thermophilic conditions. During anaerobic digestion of sewage sludge supplemented with CO added through a hollow fiber membrane (HFM) module in continuous thermophilic reactors, CO did not inhibit the process even at a pressure as high as 1.58 atm inside the HFM, due to the low dissolved CO concentration in the liquid. Complete consumption of CO was achieved with CO gas retention time of 0.2 d. Results from high-throughput sequencing analysis showed clear differences of the microbial community structures between the samples from liquid and biofilm on the HFM in the reactor with CO addition. Species close to Methanosarcina barkeri and Methanothermobacter thermautotrophicus were the two main archaeal species involved in CO biomethanation. However, the two species were distributed differently in the liquid phase and in the biofilm. Although the carboxidotrophic activities test showed that CO was converted by both archaea and bacteria, the bacterial species responsible for CO conversion are unknown.

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.

Neutron Resonance Theory for Nuclear Reactor Applications: Modern Theory and Practices.

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

Hwang, Richard N.; Blomquist, Roger N.; Leal, Luiz C.

2016-09-24

The neutron resonance phenomena constitute one of the most fundamental subjects in nuclear physics as well as in reactor physics. It is the area where the concepts of nuclear interaction and the treatment of the neutronic balance in reactor fuel lattices become intertwined. The latter requires the detailed knowledge of resonance structures of many nuclides of practical interest to the development of nuclear energy. The most essential element in reactor physics is to provide an accurate account of the intricate balance between the neutrons produced by the fission process and neutrons lost due to the absorption process as well asmore » those leaking out of the reactor system. The presence of resonance structures in many major nuclides obviously plays an important role in such processes. There has been a great deal of theoretical and practical interest in resonance reactions since Fermi’s discovery of resonance absorption of neutrons as they were slowed down in water. The resonance absorption became the center of attention when the question was raised as to the feasibility of the self-sustaining chain reaction in a natural uranium-fueled system. The threshold of the nuclear era was crossed almost eighty years ago when Fermi and Szilard observed that a substantial reduction in resonance absorption is possible if the uranium was made into the form of lumps instead of a homogeneous mixture with water. In the West, the first practical method for estimating the resonance escape probability in a reactor cell was pioneered by Wigner et al in early forties.« less

Design of single-winding energy-storage reactors for dc-to-dc converters using air-gapped magnetic-core structures

NASA Technical Reports Server (NTRS)

Ohri, A. K.; Wilson, T. G.; Owen, H. A., Jr.

1977-01-01

A procedure is presented for designing air-gapped energy-storage reactors for nine different dc-to-dc converters resulting from combinations of three single-winding power stages for voltage stepup, current stepup and voltage stepup/current stepup and three controllers with control laws that impose constant-frequency, constant transistor on-time and constant transistor off-time operation. The analysis, based on the energy-transfer requirement of the reactor, leads to a simple relationship for the required minimum volume of the air gap. Determination of this minimum air gap volume then permits the selection of either an air gap or a cross-sectional core area. Having picked one parameter, the minimum value of the other immediately leads to selection of the physical magnetic structure. Other analytically derived equations are used to obtain values for the required turns, the inductance, and the maximum rms winding current. The design procedure is applicable to a wide range of magnetic material characteristics and physical configurations for the air-gapped magnetic structure.

Structure and Activity of a New Low Molecular Weight Heparin Produced by Enzymatic Ultrafiltration

PubMed Central

FU, LI; ZHANG, FUMING; LI, GUOYUN; ONISHI, AKIHIRO; BHASKAR, UJJWAL; SUN, PEILONG; LINHARDT, ROBERT J.

2014-01-01

The standard process for preparing the low molecular weight heparin (LMWH) tinzaparin, through the partial enzymatic depolymerization of heparin, results in a reduced yield due to the formation of a high content of undesired disaccharides and tetrasaccharides. An enzymatic ultrafiltration reactor for LMWH preparation was developed to overcome this problem. The behavior, of the heparin oligosaccharides and polysaccharides using various membranes and conditions, was investigated to optimize this reactor. A novel product, LMWH-II, was produced from the controlled depolymerization of heparin using heparin lyase II in this optimized ultrafiltration reactor. Enzymatic ultrafiltration provides easy control and high yields (>80%) of LMWH-II. The molecular weight properties of LMWH-II were similar to other commercial LMWHs. The structure of LMWH-II closely matched heparin’s core structural features. Most of the common process artifacts, present in many commercial LWMHs, were eliminated as demonstrated by 1D and 2D nuclear magnetic resonance spectroscopy. The antithrombin III and platelet factor-4 binding affinity of LMWH-II were comparable to commercial LMWHs, as was its in vitro anticoagulant activity. PMID:24634007

Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

NASA Astrophysics Data System (ADS)

Natesan, K.; Li, Meimei; Chopra, O. K.; Majumdar, S.

2009-07-01

Sodium environmental effects are key limiting factors in the high temperature structural design of advanced sodium-cooled reactors. A guideline is needed to incorporate environmental effects in the ASME design rules to improve the performance reliability over long operating times. This paper summarizes the influence of sodium exposure on mechanical performance of selected austenitic stainless and ferritic/martensitic steels. Focus is on Type 316SS and mod.9Cr-1Mo. The sodium effects were evaluated by comparing the mechanical properties data in air and sodium. Carburization and decarburization were found to be the key factors that determine the tensile and creep properties of the steels. A beneficial effect of sodium exposure on fatigue life was observed under fully reversed cyclic loading in both austenitic stainless steels and ferritic/martensitic steels. However, when hold time was applied during cyclic loading, the fatigue life was significantly reduced. Based on the mechanical performance of the steels in sodium, consideration of sodium effects in high temperature structural design of advanced fast reactors is discussed.

Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

DOE PAGES

Sprouster, D. J.; Sinsheimer, J.; Dooryhee, E.; ...

2015-10-21

Here, massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that increases with dose (fluence or service time), as manifested by an increasing temperature transition from ductile-to-brittle fracture. Moreover, extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with 60 to 80 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize a new class of highly embrittling nm-scale Mn-Ni-Si precipitatesmore » that develop in the irradiated steels at high fluence. Furthermore, these precipitates can lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementarity techniques has, for the very first time, successfully characterized the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions.« less

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.

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Fielding, R.; Gan, J.

2007-09-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

Heating performances of a IC in-blanket ring array

NASA Astrophysics Data System (ADS)

Bosia, G.; Ragona, R.

2015-12-01

An important limiting factor to the use of ICRF as candidate heating method in a commercial reactor is due to the evanescence of the fast wave in vacuum and in most of the SOL layer, imposing proximity of the launching structure to the plasma boundary and causing, at the highest power level, high RF standing and DC rectified voltages at the plasma periphery, with frequent voltage breakdowns and enhanced local wall loading. In a previous work [1] the concept for an Ion Cyclotron Heating & Current Drive array (and using a different wave guide technology, a Lower Hybrid array) based on the use of periodic ring structure, integrated in the reactor blanket first wall and operating at high input power and low power density, was introduced. Based on the above concept, the heating performance of such array operating on a commercial fusion reactor is estimated.

Heating performances of a IC in-blanket ring array

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

Bosia, G., E-mail: gbosia@to.infn.it; Ragona, R.

2015-12-10

An important limiting factor to the use of ICRF as candidate heating method in a commercial reactor is due to the evanescence of the fast wave in vacuum and in most of the SOL layer, imposing proximity of the launching structure to the plasma boundary and causing, at the highest power level, high RF standing and DC rectified voltages at the plasma periphery, with frequent voltage breakdowns and enhanced local wall loading. In a previous work [1] the concept for an Ion Cyclotron Heating & Current Drive array (and using a different wave guide technology, a Lower Hybrid array) basedmore » on the use of periodic ring structure, integrated in the reactor blanket first wall and operating at high input power and low power density, was introduced. Based on the above concept, the heating performance of such array operating on a commercial fusion reactor is estimated.« less

Opportunities for Materials Science and Biological Research at the OPAL Research Reactor

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

Kennedy, S. J.

Neutron scattering techniques have evolved over more than 1/2 century into a powerful set of tools for determination of atomic and molecular structures. Modern facilities offer the possibility to determine complex structures over length scales from {approx}0.1 nm to {approx}500 nm. They can also provide information on atomic and molecular dynamics, on magnetic interactions and on the location and behaviour of hydrogen in a variety of materials. The OPAL Research Reactor is a 20 megawatt pool type reactor using low enriched uranium fuel, and cooled by water. OPAL is a multipurpose neutron factory with modern facilities for neutron beam research,more » radioisotope production and irradiation services. The neutron beam facility has been designed to compete with the best beam facilities in the world. After six years in construction, the reactor and neutron beam facilities are now being commissioned, and we will commence scientific experiments later this year. The presentation will include an outline of the strengths of neutron scattering and a description of the OPAL research reactor, with particular emphasis on it's scientific infrastructure. It will also provide an overview of the opportunities for research in materials science and biology that will be possible at OPAL, and mechanisms for accessing the facilities. The discussion will emphasize how researchers from around the world can utilize these exciting new facilities.« less

Cold adaptation and replicable microbial community development during long-term low-temperature anaerobic digestion treatment of synthetic sewage

PubMed Central

Hughes, D; Mahony, T; Cysneiros, D; Ijaz, U Z; Smith, C J; O'Flaherty, V

2018-01-01

ABSTRACT The development and activity of a cold-adapting microbial community was monitored during low-temperature anaerobic digestion (LtAD) treatment of wastewater. Two replicate hybrid anaerobic sludge bed-fixed-film reactors treated a synthetic sewage wastewater at 12°C, at organic loading rates of 0.25–1.0 kg chemical oxygen demand (COD) m−3 d−1, over 889 days. The inoculum was obtained from a full-scale anaerobic digestion reactor, which was operated at 37°C. Both LtAD reactors readily degraded the influent with COD removal efficiencies regularly exceeding 78% for both the total and soluble COD fractions. The biomass from both reactors was sampled temporally and tested for activity against hydrolytic and methanogenic substrates at 12°C and 37°C. Data indicated that significantly enhanced low-temperature hydrolytic and methanogenic activity developed in both systems. For example, the hydrolysis rate constant (k) at 12°C had increased 20–30-fold by comparison to the inoculum by day 500. Substrate affinity also increased for hydrolytic substrates at low temperature. Next generation sequencing demonstrated that a shift in a community structure occurred over the trial, involving a 1-log-fold change in 25 SEQS (OTU-free approach) from the inoculum. Microbial community structure changes and process performance were replicable in the LtAD reactors. PMID:29846574

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.

Cold adaptation and replicable microbial community development during long-term low-temperature anaerobic digestion treatment of synthetic sewage.

PubMed

Keating, C; Hughes, D; Mahony, T; Cysneiros, D; Ijaz, U Z; Smith, C J; O'Flaherty, V

2018-07-01

The development and activity of a cold-adapting microbial community was monitored during low-temperature anaerobic digestion (LtAD) treatment of wastewater. Two replicate hybrid anaerobic sludge bed-fixed-film reactors treated a synthetic sewage wastewater at 12°C, at organic loading rates of 0.25-1.0 kg chemical oxygen demand (COD) m-3 d-1, over 889 days. The inoculum was obtained from a full-scale anaerobic digestion reactor, which was operated at 37°C. Both LtAD reactors readily degraded the influent with COD removal efficiencies regularly exceeding 78% for both the total and soluble COD fractions. The biomass from both reactors was sampled temporally and tested for activity against hydrolytic and methanogenic substrates at 12°C and 37°C. Data indicated that significantly enhanced low-temperature hydrolytic and methanogenic activity developed in both systems. For example, the hydrolysis rate constant (k) at 12°C had increased 20-30-fold by comparison to the inoculum by day 500. Substrate affinity also increased for hydrolytic substrates at low temperature. Next generation sequencing demonstrated that a shift in a community structure occurred over the trial, involving a 1-log-fold change in 25 SEQS (OTU-free approach) from the inoculum. Microbial community structure changes and process performance were replicable in the LtAD reactors.

Optimization of rotational speed and hydraulic retention time of a rotational sponge reactor for sewage treatment.

PubMed

Hewawasam, Choolaka; Matsuura, Norihisa; Takimoto, Yuya; Hatamoto, Masashi; Yamaguchi, Takashi

2018-05-26

A rotational sponge (RS) reactor was proposed as an alternative sewage treatment process. Prior to the application of an RS reactor for sewage treatment, this study evaluated reactor performance with regard to organic removal, nitrification, and nitrogen removal and sought to optimize the rotational speed and hydraulic retention time (HRT) of the system. RS reactor obtained highest COD removal, nitrification, and nitrogen removal efficiencies of 91%, 97%, and 65%, respectively. For the optimization, response surface methodology (RSM) was employed and optimum conditions of rotational speed and HRT were 18 rounds per hour and 4.8 h, respectively. COD removal, nitrification, and nitrogen removal efficiencies at the optimum conditions were 85%, 85%, and 65%, respectively. Corresponding removal rates at optimum conditions were 1.6 kg-COD m -3 d -1 , 0.3 kg-NH 4 + -N m -3 d -1 , and 0.12 kg-N m -3 d -1 . Microbial community analysis revealed an abundance of nitrifying and denitrifying bacteria in the reactor, which contributed to nitrification and nitrogen removal. Copyright © 2018 Elsevier Ltd. All rights reserved.

Control rod calibration and reactivity effects at the IPEN/MB-01 reactor

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

Pinto, Letícia Negrão; Gonnelli, Eduardo; Santos, Adimir dos

2014-11-11

Researches that aim to improve the performance of neutron transport codes and quality of nuclear cross section databases are very important to increase the accuracy of simulations and the quality of the analysis and prediction of phenomena in the nuclear field. In this context, relevant experimental data such as reactivity worth measurements are needed. Control rods may be made of several neutron absorbing materials that are used to adjust the reactivity of the core. For the reactor operation, these experimental data are also extremely important: with them it is possible to estimate the reactivity worth by the movement of themore » control rod, understand the reactor response at each rod position and to operate the reactor safely. This work presents a temperature correction approach for the control rod calibration problem. It is shown the control rod calibration data of the IPEN/MB-01 reactor, the integral and differential reactivity curves and a theoretical analysis, performed by the MCNP-5 reactor physics code, developed and maintained by Los Alamos National Laboratory, using the ENDF/B-VII.0 nuclear data library.« less

FY16 Status Report for the Uranium-Molybdenum Fuel Concept

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

Bennett, Wendy D.; Doherty, Ann L.; Henager, Charles H.

2016-09-22

The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties. With sufficient development, it may be able to provide the Light Water Reactor industry with a melt-resistant, accident-tolerant fuel with improved safety response. The Pacific Northwest National Laboratory has been tasked with extrusion development and performing ex-reactor corrosion testing to characterize the performance of Uranium-Molybdenum fuel in both these areas. This report documents the results of the fiscal yearmore » 2016 effort to develop the Uranium-Molybdenum metal fuel concept for light water reactors.« less

The performance of hafnium and gadolinium self powered neutron detectors in the TREAT reactor

NASA Astrophysics Data System (ADS)

Imel, G. R.; Hart, P. R.

1996-05-01

The use of gadolinium and hafnium self powered neutron detectors in a transient reactor is described in this paper. The detectors were calibrated to the fission rate of U-235 using calibrated fission chambers; the calibration factors were tested in two reactors in steady state and found to be consistent. Calibration of the detectors in transient reactor conditions was done by using uranium wires that were analyzed by radiochemistry techniques to determine total fissions during the transient. This was correlated to the time-integrated current of the detectors during the transient. A temperature correction factor was derived to account for self-shielding effects in the hafnium and gadolinium detectors. The dynamic response of the detectors under transient conditions was studied, and found to be excellent.

R and D program for core instrumentation improvements devoted for French sodium fast reactors

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

Jeannot, J. P.; Rodriguez, G.; Jammes, C.

2011-07-01

Under the framework of French R and D studies for Generation IV reactors and more specifically for sodium-cooled fast reactors (SFR); the CEA, EDF and AREVA have launched a joint coordinated research programme. This paper deals with the R and D sets out to achieve better inspection, maintenance, availability and decommissioning. In particular the instrumentation requirements for core monitoring and detection in the case of accidental events. Requirements mainly involve diversifying the means of protection and improving instrumentation performance in terms of responsiveness and sensitivity. Operation feedback from the Phenix and Superphenix prototype reactors and studies, carried out within themore » scope of the EFR projects, has been used to define the needs for instrumentation enhancement. (authors)« less

FERRET-SAND II physics-dosimetry analysis for N Reactor Pressure Tubes 2954, 3053 and 1165 using a WIMS calculated input spectrum

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

McElroy, W.N.; Kellogg, L.S.; Matsumoto, W.Y.

1988-05-01

This report is in response to a request from Westinghouse Hanford Company (WHC) that the PNL National Dosimetry Center (NDC) perform physics-dosimetry analyses (E > MeV) for N Reactor Pressure Tubes 2954 and 3053. As a result of these analyses, and recommendations for additional studies, two physics-dosimetry re-evaluations for Pressure Tube 1165 were also accomplished. The primary objective of Pacific Northwest Laboratories' (PNL) National Dosimetry Center (NDC) physics-dosimetry work for N Reactor was to provide FERRET-SAND II physics-dosimetry results to assist in the assessment of neutron radiation-induced changes in the physical and mechanical properties of N Reactor pressure tubes. 15more » refs., 6 figs., 5 tabs.« less

Analysis by gender and Visual Imagery Reactivity of conventional and imagery Rorschach.

PubMed

Yanovski, A; Menduke, H; Albertson, M G

1995-06-01

Examined here are the effects of gender and Visual Imagery Reactivity in 80 consecutively selected psychiatric outpatients. The participants were grouped by gender and by the amounts of responsiveness to preceding therapy work using imagery (Imagery Nonreactors and Reactors). In the group of Imagery Nonreactors were 13 men and 22 women, and in the Reactor group were 17 men and 28 women. Compared were the responses to standard Rorschach (Conventional condition) with visual associations to memory images of Rorschach inkblots (Imagery condition). Responses were scored using the Visual Imagery Reactivity (VIR) scoring system, a general, test-nonspecific scoring method. Nonparametric statistical analysis showed that critical indicators of Imagery Reactivity encoded as High Affect/Conflict score and its derivatives associated with sexual or bizarre content were not significantly associated with gender; neither was Neutral Content score which categorizes "non-Reactivity." These results support the notion that system's criteria of Visual Imagery Reactivity can be applied equally to both men and women for the classification of Imagery Reactors and Nonreactors. Discussed are also the speculative consequences of extending the tolerance range of significance levels for the interaction between Reactivity and sex above the customary limit of p < .05 in borderline cases. The results of such an analysis may imply a trend towards more rigid defensiveness under Imagery and toward lesser verbal productivity in response to either the Conventional or the Imagery task among women who are Nonreactors. In Reactors, men produced significantly more Sexual Reference scores (in the subcategory not associated with High Affect/Conflict) than women, but this could be attributed to the effect of tester's and subjects' gender combined.

Neutronic Reactor III

NASA Astrophysics Data System (ADS)

Fermi, Enrico; Zinn, Walter H.; Anderson, Herbert L.

An improvement of the reactors described in the previous Patents, aimed at increasing the reproduction factor, is reported here, such improvement being obtained by diminishing the neutron loss due to impurities within the reactor. This is achieved by encasing the reactor in a rubberized balloon cloth housing (or something like this) in order to eliminate the atmospheric air therefrom, thus eliminating both the effect of the danger coefficient of nitrogen (70% of the atmospheric air) and that of the argon present in the air, which can become radioactive. Since the removal of the air from the reactor may result in structural problems, caused by the forces brought into play by that evacuation, the reactor is then filled with a non-reactive (from a chemical and nuclear standpoint) gas such as helium or carbon dioxide. It is interesting to point out that the authors consider also the possibility to control (a little) the reproduction ratio of the reactor by varying the air content of it. Just a rapid mention of the main idea of the present Patent (i.e. the encasing of the pile in a balloon cloth) appeared in [Fermi (1942f)], but no detailed description of the system considered here is reported in any other published paper.

Emulation of reactor irradiation damage using ion beams

DOE PAGES

Was, G. S.; Jiao, Z.; Getto, E.; ...

2014-06-14

The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide,more » irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.« less

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.

Experiences in utilization of research reactors in Yugoslavia

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

Copic, M.; Gabrovsek, Z.; Pop-Jordanov, J.

1971-06-15

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied bymore » means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro envisage the role of research reactors in the promotion of nuclear power programs in relation to the size of the program, the competence of domestic industries and the degree of independence where fuel supply is concerned. (author)« less

Studies of neutron-rich nuclei far from stability at TRISTAN

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

Gill, R.L.

The ISOL facility, TRISTAN, is a user facility located at Brookhaven National Laboratory's High Flux Beam Reactor. Short-lived, neutron-rich nuclei, far from stability, are produced by thermal neutron fission of /sup 235/U. An extensive array of experimental end stations are available for nuclear structure studies. These studies are augmented by a variety of long-lived ion sources suitable for use at a reactor facility. Some recent results at TRISTAN are presented as examples of using an ISOL facility to study series of nuclei, whereby an effective means of conducting nuclear structure investigations is available.

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

Bearinger, J P

This months issue has the following articles: (1) Science Translated for the Greater Good--Commentary by Steven D. Liedle; (2) The New Face of Industrial Partnerships--An entrepreneurial spirit is blossoming at Lawrence Livermore; (3) Monitoring a Nuclear Weapon from the Inside--Livermore researchers are developing tiny sensors to warn of detrimental chemical and physical changes inside nuclear warheads; (4) Simulating the Biomolecular Structure of Nanometer-Size Particles--Grand Challenge simulations reveal the size and structure of nanolipoprotein particles used to study membrane proteins; and (5) Antineutrino Detectors Improve Reactor Safeguards--Antineutrino detectors track the consumption and production of fissile materials inside nuclear reactors.

Grain boundary engineering for structure materials of nuclear reactors

NASA Astrophysics Data System (ADS)

Tan, L.; Allen, T. R.; Busby, J. T.

2013-10-01

Grain boundary engineering (GBE), primarily implemented by thermomechanical processing, is an effective and economical method of enhancing the properties of polycrystalline materials. Among the factors affecting grain boundary character distribution, literature data showed definitive effect of grain size and texture. GBE is more effective for austenitic stainless steels and Ni-base alloys compared to other structural materials of nuclear reactors, such as refractory metals, ferritic and ferritic-martensitic steels, and Zr alloys. GBE has shown beneficial effects on improving the strength, creep strength, and resistance to stress corrosion cracking and oxidation of austenitic stainless steels and Ni-base alloys.

Shock and vibration tests of uranium mononitride fuel pellets for a space power nuclear reactor

NASA Technical Reports Server (NTRS)

Adams, D. W.

1972-01-01

Shock and vibration tests were conducted on cylindrically shaped, depleted, uranium mononitride (UN) fuel pellets. The structural capabilities of the pellets were determined under exposure to shock and vibration loading which a nuclear reactor may encounter during launching into space. Various combinations of diametral and axial clearances between the pellets and their enclosing structures were tested. The results of these tests indicate that for present fabrication of UN pellets, a diametral clearance of 0.254 millimeter and an axial clearance of 0.025 millimeter are tolerable when subjected to launch-induced loads.

Structural behavior of the Bitter plate tf magnet for the Zephyr ignition test reactor

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

Bobrov, E.S.; Becker, H.

1981-01-01

This paper discusses methods and results of the computer structural analysis of the Bitter plate toroidal field magnet design for the ZEPHYR Ignition Test Reactor. The magnet provides a field of 7.06 T at the center of the bore which is 1.76 m from the major toroidal axis. The ignited plasma is located at a major radius of 1.36 m where the magnetic field is 9.11 T. The plasma is moved to this final position following compression in the major radius. The horizontal bore of the magnet is 1.8 m.

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.

A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

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

John Darrell Bess

2008-06-01

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, suchmore » as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces.« less

The use of experimental data in an MTR-type nuclear reactor safety analysis

NASA Astrophysics Data System (ADS)

Day, Simon E.

Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

Space nuclear reactors — A post-operational disposal strategy

NASA Astrophysics Data System (ADS)

Angelo, Joseph A.; Buden, David

If 100-kWe and multimegawatt-electric class space nuclear reactors are to play a significant role in humanity's push into cislunar and heliocentric space in the next millennium, the obvious advantages of space nuclear power plants should not be denied to space mission planners due to a failure to develop internationally-acceptable post-operational disposal strategies for spent reactor cores. This is true whether the space reactor has shut down at the end of its normal mission lifetime or in response to an onboard system failure/emergency which causes a premature mission termination. Up until now the great majority of aerospace nuclear safety efforts have concentrated on prelaunch, launch and reactor startup activities. In fact, with the exception of the development of the "nuclear safe orbit" (NSO) concept, little technical attention has yet been given to the post-operational disposal of future space reactors. This paper describes the technical alternatives available for the safe, acceptable disposal of space reactors that could be used in a wide variety of space applications in the 21st Century. Post-operational core radioactivity levels for typical advanced design (hundred kWe-class) space reactors are presented as a function of decay time and contrasted to the spent core radionuclide inventory of the SNAP-10A system, the only nuclear reactor operated in space by the United States. The role of a permanent space station, smart robotic systems, and an operating lunar base in support of spent core disposal strategies is also presented, including use of a selected portion of the lunar surface as an internationally-designated spent reactor core repository.

Design feasibility study of a divertor component reinforced with fibrous metal matrix composite laminate

NASA Astrophysics Data System (ADS)

You, Jeong-Ha

2005-01-01

Fibrous metal matrix composites possess advanced mechanical properties compared to conventional alloys. It is expected that the application of these composites to a divertor component will enhance the structural reliability. A possible design concept would be a system consisting of tungsten armour, copper composite interlayer and copper heat sink where the composite interlayer is locally inserted into the highly stressed domain near the bond interface. For assessment of the design feasibility of the composite divertor concept, a non-linear multi-scale finite element analysis was performed. To this end, a micro-mechanics algorithm was implemented into a finite element code. A reactor-relevant heat flux load was assumed. Focus was placed on the evolution of stress state, plastic deformation and ductile damage on both macro- and microscopic scales. The structural response of the component and the micro-scale stress evolution of the composite laminate were investigated.

Exploring the engineering limit of heat flux of a W/RAFM divertor target for fusion reactors

NASA Astrophysics Data System (ADS)

Mao, X.; Fursdon, M.; Chang, X. B.; Zhang, J. W.; Liu, P.; Ellwood, G.; Qian, X. Y.; Qin, S. J.; Peng, X. B.; Barrett, T. R.; Liu, P.

2018-06-01

The design and development of a fusion reactor divertor plasma facing component (PFC) is one of the many challenging issues on the road to commercial use of fusion energy. The divertor PFC is expected to exhaust steady state heat loads in the region of 10 MW m‑2 while keeping temperatures and thermo-mechanical stresses in its structure within the allowable limits. For ITER (International Thermo-Nuclear Experimental Reactor) a water cooled W/CuCrZr divertor PFC concept has been developed. However, this concept is not necessarily assured for use in future fusion reactors mainly because the neutron radiation dose would be at least an order magnitude higher, resulting in limited thermo-mechanical performance and considerably more activated waste products. In the present study, a water cooled divertor PFC using reduced activation ferritic-martensitic (RAFM) steel as the heat sink pipe has been designed with pressurised water reactor-like cooling conditions (pressure of 15.5 MPa, velocity of 10–20 m s‑1 and temperature of 300 °C). The PFC is made up of a number of rectangular tungsten tiles, each with an inner circular hole (so-called monoblocks), joined onto a RAFM steel pipe with copper interlayers. The thermo-mechanical performance of the PFC has been studied in detail. The heat transfer coefficient between the RAFM pipe inner surface and the water was calculated using published correlations. Geometric parameters and water velocity were optimized with finite element (FE) thermal analysis, to achieve acceptable temperatures in the structure given the target exhaust heat load of 10 MW m‑2. Under this heat load and the optimised thermal design parameters, the structure of the PFC was further assessed by mechanical analysis. We find that under these conditions the RAFM steel pipe experiences cyclic plasticity, and fails the common linear elastic ratchetting (3 Sm) rule. Nevertheless, the designed W/RAFM divertor PFU can withstand 10 MW m‑2 heat load, albeit with a fatigue life of approximately 0.55 years based on the expected operation scenario of a prototype or test reactor. This study extends the state of knowledge of the technological limit of a divertor based on a RAFM steel pipe structure.

NEUTRONIC REACTOR

DOEpatents

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

1958-04-15

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

NEUTRONIC REACTOR SHIELD AND SPACER CONSTRUCTION

DOEpatents

Wigner, E.P.; Ohlinger, L.A.

1958-11-18

Reactors of the heterogeneous, graphite moderated, fluid cooled type and shielding and spacing plugs for the coolant channels thereof are reported. In this design, the coolant passages extend horizontally through the moderator structure, accommodating the fuel elements in abutting end-to-end relationship, and have access openings through the outer shield at one face of the reactor to facilitate loading of the fuel elements. In the outer ends of the channels which extend through the shields are provided spacers and shielding plugs designed to offer minimal reslstance to coolant fluid flow while preventing emanation of harmful radiation through the access openings when closed between loadings.

Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

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

Tsai, H.; Gomes, I.C.; Smith, D.L.

1998-09-01

The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

Vibration monitoring of Kraftwerk Union pressurized water reactors - Review, present status, and further development

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

Stolben, H.; Wehling, H.J.

Incipient damage to mechanical structure may be detected early in time by deviations from normal dynamic behavior. For vibration monitoring of coupled systems, only a small number of transducers are necessary, in general. On the basis, Kraftwerk Union has been involved in the development and construction of vibration monitoring systems for pressurized water reactors over the last 20 yr. The current state of the art permits vibration monitoring during normal operation by reactor personnel without expert assistance. The new SUS-86 microprocessor-based system allows further expansion toward an expert system.

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

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

None

2013-09-25

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

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

ScienceCinema

None

2018-01-16

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

Support vector regression model of wastewater bioreactor performance using microbial community diversity indices: effect of stress and bioaugmentation.

PubMed

Seshan, Hari; Goyal, Manish K; Falk, Michael W; Wuertz, Stefan

2014-04-15

The relationship between microbial community structure and function has been examined in detail in natural and engineered environments, but little work has been done on using microbial community information to predict function. We processed microbial community and operational data from controlled experiments with bench-scale bioreactor systems to predict reactor process performance. Four membrane-operated sequencing batch reactors treating synthetic wastewater were operated in two experiments to test the effects of (i) the toxic compound 3-chloroaniline (3-CA) and (ii) bioaugmentation targeting 3-CA degradation, on the sludge microbial community in the reactors. In the first experiment, two reactors were treated with 3-CA and two reactors were operated as controls without 3-CA input. In the second experiment, all four reactors were additionally bioaugmented with a Pseudomonas putida strain carrying a plasmid with a portion of the pathway for 3-CA degradation. Molecular data were generated from terminal restriction fragment length polymorphism (T-RFLP) analysis targeting the 16S rRNA and amoA genes from the sludge community. The electropherograms resulting from these T-RFs were used to calculate diversity indices - community richness, dynamics and evenness - for the domain Bacteria as well as for ammonia-oxidizing bacteria in each reactor over time. These diversity indices were then used to train and test a support vector regression (SVR) model to predict reactor performance based on input microbial community indices and operational data. Considering the diversity indices over time and across replicate reactors as discrete values, it was found that, although bioaugmentation with a bacterial strain harboring a subset of genes involved in the degradation of 3-CA did not bring about 3-CA degradation, it significantly affected the community as measured through all three diversity indices in both the general bacterial community and the ammonia-oxidizer community (α = 0.5). The impact of bioaugmentation was also seen qualitatively in the variation of community richness and evenness over time in each reactor, with overall community richness falling in the case of bioaugmented reactors subjected to 3-CA and community evenness remaining lower and more stable in the bioaugmented reactors as opposed to the unbioaugmented reactors. Using diversity indices, 3-CA input, bioaugmentation and time as input variables, the SVR model successfully predicted reactor performance in terms of the removal of broad-range contaminants like COD, ammonia and nitrate as well as specific contaminants like 3-CA. This work was the first to demonstrate that (i) bioaugmentation, even when unsuccessful, can produce a change in community structure and (ii) microbial community information can be used to reliably predict process performance. However, T-RFLP may not result in the most accurate representation of the microbial community itself, and a much more powerful prediction tool can potentially be developed using more sophisticated molecular methods. Copyright © 2014 Elsevier Ltd. All rights reserved.

Flow duct for nuclear reactors

DOEpatents

Straalsund, Jerry L.

1978-01-01

Improved liquid sodium flow ducts for nuclear reactors are described wherein the improvement comprises varying the wall thickness of each of the walls of a polygonal tubular duct structure so that each of the walls is of reduced cross-section along the longitudinal center line and of a greater cross-section along wall junctions with the other walls to form the polygonal tubular configuration.

Light-Water-Reactor safety research program. Quarterly progress report, January--March 1977

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

None

The report summarizes the Argonne National Laboratory work performed during January, February, and March 1977 on water-reactor-safety problems. The following research and development areas are covered: (1) loss-of-coolant accident research: heat transfer and fluid dynamics; (2) transient fuel response and fission-product release program; (3) mechanical properties of zircaloy containing oxygen; and (4) steam-explosion studies.

After Action Report: Advanced Test Reactor Complex 2015 Evaluated Drill October 6, 2015

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

Holmes, Forest Howard

2015-11-01

The Advanced Test Reactor (ATR) Complex, operated by Battelle Energy Alliance, LLC, at the Idaho National Laboratory (INL) conducted an evaluated drill on October 6, 2015, to allow the ATR Complex emergency response organization (ERO) to demonstrate the ability to respond to and mitigate an emergency by implementing the requirements of DOE O 151.1C, “Comprehensive Emergency Management System.”

Early Program Development

NASA Image and Video Library

2004-04-15

This artist's concept illustrates the NERVA (Nuclear Engine for Rocket Vehicle Application) engine's hot bleed cycle in which a small amount of hydrogen gas is diverted from the thrust nozzle, thus eliminating the need for a separate system to drive the turbine. The NERVA engine, based on KIWI nuclear reactor technology, would power a RIFT (Reactor-In-Flight-Test) nuclear stage, for which the Marshall Space Flight Center had development responsibility.

Transport Corrections in Nodal Diffusion Codes for HTR Modeling

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

Abderrafi M. Ougouag; Frederick N. Gleicher

2010-08-01

The cores and reflectors of High Temperature Reactors (HTRs) of the Next Generation Nuclear Plant (NGNP) type are dominantly diffusive media from the point of view of behavior of the neutrons and their migration between the various structures of the reactor. This means that neutron diffusion theory is sufficient for modeling most features of such reactors and transport theory may not be needed for most applications. Of course, the above statement assumes the availability of homogenized diffusion theory data. The statement is true for most situations but not all. Two features of NGNP-type HTRs require that the diffusion theory-based solutionmore » be corrected for local transport effects. These two cases are the treatment of burnable poisons (BP) in the case of the prismatic block reactors and, for both pebble bed reactor (PBR) and prismatic block reactor (PMR) designs, that of control rods (CR) embedded in non-multiplying regions near the interface between fueled zones and said non-multiplying zones. The need for transport correction arises because diffusion theory-based solutions appear not to provide sufficient fidelity in these situations.« less

Radiation chemistry for modern nuclear energy development

NASA Astrophysics Data System (ADS)

Chmielewski, Andrzej G.; Szołucha, Monika M.

2016-07-01

Radiation chemistry plays a significant role in modern nuclear energy development. Pioneering research in nuclear science, for example the development of generation IV nuclear reactors, cannot be pursued without chemical solutions. Present issues related to light water reactors concern radiolysis of water in the primary circuit; long-term storage of spent nuclear fuel; radiation effects on cables and wire insulation, and on ion exchangers used for water purification; as well as the procedures of radioactive waste reprocessing and storage. Radiation effects on materials and enhanced corrosion are crucial in current (II/III/III+) and future (IV) generation reactors, and in waste management, deep geological disposal and spent fuel reprocessing. The new generation of reactors (III+ and IV) impose new challenges for radiation chemists due to their new conditions of operation and the usage of new types of coolant. In the case of the supercritical water-cooled reactor (SCWR), water chemistry control may be the key factor in preventing corrosion of reactor structural materials. This paper mainly focuses on radiation effects on long-term performance and safety in the development of nuclear power plants.

Start-up performance and granular sludge features of an improved external circulating anaerobic reactor for algae-laden water treatment.

PubMed

Yu, Yaqin; Lu, Xiwu

2017-09-01

The microbial characteristics of granular sludge during the rapid start of an enhanced external circulating anaerobic reactor were studied to improve algae-laden water treatment efficiency. Results showed that algae laden water was effectively removed after about 35 d, and the removal rates of chemical oxygen demand (COD) and algal toxin were around 85% and 92%, respectively. Simultaneously, the gas generation rate was around 380 mL/gCOD. The microbial community structure in the granular sludge of the reactor was complicated, and dominated by coccus and filamentous bacteria. Methanosphaera , Methanolinea , Thermogymnomonas , Methanoregula , Methanomethylovorans , and Methanosaeta were the major microorganisms in the granular sludge. The activities of protease and coenzyme F 420 were high in the granular sludge. The intermittent stirring device and the reverse-flow system were further found to overcome the disadvantage of the floating and crusting of cyanobacteria inside the reactor. Meanwhile, the effect of mass transfer inside the reactor can be accelerated to help give the reactor a rapid start.

A highly efficient autothermal microchannel reactor for ammonia decomposition: Analysis of hydrogen production in transient and steady-state regimes

NASA Astrophysics Data System (ADS)

Engelbrecht, Nicolaas; Chiuta, Steven; Bessarabov, Dmitri G.

2018-05-01

The experimental evaluation of an autothermal microchannel reactor for H2 production from NH3 decomposition is described. The reactor design incorporates an autothermal approach, with added NH3 oxidation, for coupled heat supply to the endothermic decomposition reaction. An alternating catalytic plate arrangement is used to accomplish this thermal coupling in a cocurrent flow strategy. Detailed analysis of the transient operating regime associated with reactor start-up and steady-state results is presented. The effects of operating parameters on reactor performance are investigated, specifically, the NH3 decomposition flow rate, NH3 oxidation flow rate, and fuel-oxygen equivalence ratio. Overall, the reactor exhibits rapid response time during start-up; within 60 min, H2 production is approximately 95% of steady-state values. The recommended operating point for steady-state H2 production corresponds to an NH3 decomposition flow rate of 6 NL min-1, NH3 oxidation flow rate of 4 NL min-1, and fuel-oxygen equivalence ratio of 1.4. Under these flows, NH3 conversion of 99.8% and H2 equivalent fuel cell power output of 0.71 kWe is achieved. The reactor shows good heat utilization with a thermal efficiency of 75.9%. An efficient autothermal reactor design is therefore demonstrated, which may be upscaled to a multi-kW H2 production system for commercial implementation.

Anaerobic biodegradation of aircraft deicing fluid in UASB reactors.

PubMed

Tham, P T Pham thi; Kennedy, K J Kevin J

2004-05-01

A central composite design was employed to methodically investigate anaerobic treatment of aircraft deicing fluid (ADF) in bench-scale Upflow Anaerobic Sludge Blanket (UASB) reactors. A total of 23 runs at 17 different operating conditions (0.8% 1.6% ADF (6000-12,000mg/L COD), 12-56h HRT, and 18-36gVSS/L) were conducted in continuous mode. The development of four empirical models describing process responses (i.e. COD removal efficiency, biomass-specific acetoclastic activity, methane production rate, and methane production potential) as functions of ADF concentration, hydraulic retention time, and biomass concentration is presented. Model verification indicated that predicted responses (COD removal efficiencies, biomass-specific acetoclastic activity, and methane production rates and potential) were in good agreement with experimental results. Biomass-specific acetoclastic activity was improved two-fold from 0.23gCOD/gVSS/d for inoculum to a maximum of 0.55gCOD/gVSS/d during ADF treatment in UASB reactors. For the design window, COD removal efficiencies were higher than 90%. The predicted methane production potentials were close to theoretical values, and methane production rates increased as the organic loading rate is increased. ADF toxicity effects were evident for 1.6% ADF at medium organic loadings (SOLR above 0.5gCOD/gVSS/d). In contrast, good reactor stability and excellent COD removal efficiencies were achieved at 1.2% ADF for reactor loadings approaching that of highly loaded systems (0.73gCOD/gVSS/d).

Community proteomics provides functional insight into polyhydroxyalkanoate production by a mixed microbial culture cultivated on fermented dairy manure.

PubMed

Hanson, Andrea J; Guho, Nicholas M; Paszczynski, Andrzej J; Coats, Erik R

2016-09-01

Polyhydroxyalkanoates (PHAs) are bio-based, biodegradable polyesters that can be produced from organic-rich waste streams using mixed microbial cultures (MMCs). To maximize PHA production, MMCs are enriched for bacteria with a high polymer storage capacity through the application of aerobic dynamic feeding (ADF) in a sequencing batch reactor (SBR), which consequently induces a feast-famine metabolic response. Though the feast-famine response is generally understood empirically at a macro-level, the molecular level is less refined. The objective of this study was to investigate the microbial community composition and proteome profile of an enriched MMC cultivated on fermented dairy manure. The enriched MMC exhibited a feast-famine response and was capable of producing up to 40 % (wt. basis) PHA in a fed-batch reactor. High-throughput 16S rRNA gene sequencing revealed a microbial community dominated by Meganema, a known PHA-producing genus not often observed in high abundance in enrichment SBRs. The application of the proteomic methods two-dimensional electrophoresis and LC-MS/MS revealed PHA synthesis, energy generation, and protein synthesis prominently occurring during the feast phase, corroborating bulk solution variable observations and theoretical expectations. During the famine phase, nutrient transport, acyl-CoA metabolism, additional energy generation, and housekeeping functions were more pronounced, informing previously under-determined MMC functionality under famine conditions. During fed-batch PHA production, acetyl-CoA acetyltransferase and PHA granule-bound phasin proteins were in increased abundance relative to the SBR, supporting the higher PHA content observed. Collectively, the results provide unique microbial community structural and functional insight into feast-famine PHA production from waste feedstocks using MMCs.

Achieving nitritation in a continuous moving bed biofilm reactor at different temperatures through ratio control.

PubMed

Bian, Wei; Zhang, Shuyan; Zhang, Yanzhuo; Li, Wenjing; Kan, Ruizhe; Wang, Wenxiao; Zheng, Zhaoming; Li, Jun

2017-02-01

A ratio control strategy was implemented in a continuous moving bed biofilm reactor (MBBR) to investigate the response to different temperatures. The control strategy was designed to maintain a constant ratio between dissolved oxygen (DO) and total ammonia nitrogen (TAN) concentrations. The results revealed that a stable nitritation in a biofilm reactor could be achieved via ratio control, which compensated the negative influence of low temperatures by stronger oxygen-limiting conditions. Even with a temperature as low as 6°C, stable nitritation could be achieved when the controlling ratio did not exceed 0.17. Oxygen-limiting conditions in the biofilm reactor were determined by the DO/TAN concentrations ratio, instead of the mere DO concentration. This ratio control strategy allowed the achievement of stable nitritation without complete wash-out of NOB from the reactor. Through the ratio control strategy full nitritation of sidestream wastewater was allowed; however, for mainstream wastewater, only partial nitritation was recommended. Copyright © 2016 Elsevier Ltd. All rights reserved.

Multi-objective optimization of oxidative desulfurization in a sono-photochemical airlift reactor.

PubMed

Behin, Jamshid; Farhadian, Negin

2017-09-01

Response surface methodology (RSM) was employed to optimize ultrasound/ultraviolet-assisted oxidative desulfurization in an airlift reactor. Ultrasonic waves were incorporated in a novel-geometry reactor to investigate the synergistic effects of sono-chemistry and enhanced gas-liquid mass transfer. Non-hydrotreated kerosene containing sulfur and aromatic compounds was chosen as a case study. Experimental runs were conducted based on a face-centered central composite design and analyzed using RSM. The effects of two categorical factors, i.e., ultrasound and ultraviolet irradiation and two numerical factors, i.e., superficial gas velocity and oxidation time were investigated on two responses, i.e., desulfurization and de-aromatization yields. Two-factor interaction (2FI) polynomial model was developed for the responses and the desirability function associate with overlay graphs was applied to find optimum conditions. The results showed enhancement in desulfurization ability corresponds to more reduction in aromatic content of kerosene in each combination. Based on desirability approach and certain criteria considered for desulfurization/de-aromatization, the optimal desulfurization and de-aromatization yields of 91.7% and 48% were obtained in US/UV/O 3 /H 2 O 2 combination, respectively. Copyright © 2017 Elsevier B.V. All rights reserved.

Development of an In-Situ Decommissioning Sensor Network Test Bed for Structural Condition Monitoring - 12156

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

Zeigler, Kristine E.; Ferguson, Blythe A.

2012-07-01

The Savannah River National Laboratory (SRNL) has established an In Situ Decommissioning (ISD) Sensor Network Test Bed, a unique, small scale, configurable environment, for the assessment of prospective sensors on actual ISD system material, at minimal cost. The Department of Energy (DOE) is presently implementing permanent entombment of contaminated, large nuclear structures via ISD. The ISD end state consists of a grout-filled concrete civil structure within the concrete frame of the original building. Validation of ISD system performance models and verification of actual system conditions can be achieved through the development a system of sensors to monitor the materials andmore » condition of the structure. The ISD Sensor Network Test Bed has been designed and deployed to addresses the DOE-Environmental Management Technology Need to develop a remote monitoring system to determine and verify ISD system performance. Commercial off-the-shelf sensors have been installed on concrete blocks taken from walls of the P Reactor Building at the Savannah River Site. Deployment of this low-cost structural monitoring system provides hands-on experience with sensor networks. The initial sensor system consists of groutable thermistors for temperature and moisture monitoring, strain gauges for crack growth monitoring, tilt-meters for settlement monitoring, and a communication system for data collection. Baseline data and lessons learned from system design and installation and initial field testing will be utilized for future ISD sensor network development and deployment. The Sensor Network Test Bed at SRNL uses COTS sensors on concrete blocks from the outer wall of the P Reactor Building to measure conditions expected to occur in ISD structures. Knowledge and lessons learned gained from installation, testing, and monitoring of the equipment will be applied to sensor installation in a meso-scale test bed at FIU and in future ISD structures. The initial data collected from the sensors installed on the P Reactor Building blocks define the baseline materials condition of the P Reactor ISD external concrete structure. Continued monitoring of the blocks will enable evaluation of the effects of aging on the P Reactor ISD structure. The collected data will support validation of the material degradation model and assessment of the condition of the ISD structure over time. The following are recommendations for continued development of the ISD Sensor Network Test Bed: - Establish a long-term monitoring program using the concrete blocks with existing sensor and/or additional sensors for trending the concrete materials and structural condition; - Continue development of a stand-alone test bed sensor system that is self-powered and provides wireless transmission of data to a user-accessible dashboard; - Develop and implement periodic NDE/DE characterization of the concrete blocks to provide verification and validation for the measurements obtained through the sensor system and concrete degradation model(s). (authors)« less

Turbulence accelerates the growth of drinking water biofilms.

PubMed

Tsagkari, E; Sloan, W T

2018-06-01

Biofilms are found at the inner surfaces of drinking water pipes and, therefore, it is essential to understand biofilm processes to control their formation. Hydrodynamics play a crucial role in shaping biofilms. Thus, knowing how biofilms form, develop and disperse under different flow conditions is critical in the successful management of these systems. Here, the development of biofilms after 4 weeks, the initial formation of biofilms within 10 h and finally, the response of already established biofilms within 24-h intervals in which the flow regime was changed, were studied using a rotating annular reactor under three different flow regimes: turbulent, transition and laminar. Using fluorescence microscopy, information about the number of microcolonies on the reactor slides, the surface area of biofilms and of extracellular polymeric substances and the biofilm structures was acquired. Gravimetric measurements were conducted to characterise the thickness and density of biofilms, and spatial statistics were used to characterise the heterogeneity and spatial correlation of biofilm structures. Contrary to the prevailing view, it was shown that turbulent flow did not correlate with a reduction in biofilms; turbulence was found to enhance both the initial formation and the development of biofilms on the accessible surfaces. Additionally, after 24-h changes of the flow regime it was indicated that biofilms responded to the quick changes of the flow regime. Overall, this work suggests that different flow conditions can cause substantial changes in biofilm morphology and growth and specifically that turbulent flow can accelerate biofilm growth in drinking water.

Deterministic Local Sensitivity Analysis of Augmented Systems - II: Applications to the QUENCH-04 Experiment Using the RELAP5/MOD3.2 Code System

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

Ionescu-Bujor, Mihaela; Jin Xuezhou; Cacuci, Dan G.

2005-09-15

The adjoint sensitivity analysis procedure for augmented systems for application to the RELAP5/MOD3.2 code system is illustrated. Specifically, the adjoint sensitivity model corresponding to the heat structure models in RELAP5/MOD3.2 is derived and subsequently augmented to the two-fluid adjoint sensitivity model (ASM-REL/TF). The end product, called ASM-REL/TFH, comprises the complete adjoint sensitivity model for the coupled fluid dynamics/heat structure packages of the large-scale simulation code RELAP5/MOD3.2. The ASM-REL/TFH model is validated by computing sensitivities to the initial conditions for various time-dependent temperatures in the test bundle of the Quench-04 reactor safety experiment. This experiment simulates the reflooding with water ofmore » uncovered, degraded fuel rods, clad with material (Zircaloy-4) that has the same composition and size as that used in typical pressurized water reactors. The most important response for the Quench-04 experiment is the time evolution of the cladding temperature of heated fuel rods. The ASM-REL/TFH model is subsequently used to perform an illustrative sensitivity analysis of this and other time-dependent temperatures within the bundle. The results computed by using the augmented adjoint sensitivity system, ASM-REL/TFH, highlight the reliability, efficiency, and usefulness of the adjoint sensitivity analysis procedure for computing time-dependent sensitivities.« less

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

NASA Astrophysics Data System (ADS)

Qvist, Staffan Alexander

In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

Bacterial community structure in experimental methanogenic bioreactors and search for pathogenic clostridia as community members.

PubMed

Dohrmann, Anja B; Baumert, Susann; Klingebiel, Lars; Weiland, Peter; Tebbe, Christoph C

2011-03-01

Microbial conversion of organic waste or harvested plant material into biogas has become an attractive technology for energy production. Biogas is produced in reactors under anaerobic conditions by a consortium of microorganisms which commonly include bacteria of the genus Clostridium. Since the genus Clostridium also harbors some highly pathogenic members in its phylogenetic cluster I, there has been some concern that an unintended growth of such pathogens might occur during the fermentation process. Therefore this study aimed to follow how process parameters affect the diversity of Bacteria in general, and the diversity of Clostridium cluster I members in particular. The development of both communities was followed in model biogas reactors from start-up during stable methanogenic conditions. The biogas reactors were run with either cattle or pig manures as substrates, and both were operated at mesophilic and thermophilic conditions. The structural diversity was analyzed independent of cultivation using a PCR-based detection of 16S rRNA genes and genetic profiling by single-strand conformation polymorphism (SSCP). Genetic profiles indicated that both bacterial and clostridial communities evolved in parallel, and the community structures were highly influenced by both substrate and temperature. Sequence analysis of 16S rRNA genes recovered from prominent bands from SSCP profiles representing Clostridia detected no pathogenic species. Thus, this study gave no indication that pathogenic clostridia would be enriched as dominant community members in biogas reactors fed with manure.

Radiation Effects in Fission and Fusion Reactors

NASA Astrophysics Data System (ADS)

Odette, G. Robert; Wirth, Brian D.

Since the prediction of "Wigner disease" [1] and the subsequent observation of anisotropic growth of the graphite used in the Chicago Pile, the effects of radiation on materials has been an important technological concern. The broad field of radiation effects impacts many critical advanced technologies, ranging from semiconductor processing to severe materials degradation in nuclear reactor environments. Radiation effects also occur in many natural environments, ranging from deep space to inside the Earth's crust. As selected examples that involve many basic phenomena that cross-cut and illustrate the broader impacts of radiation exposure on materials, this article focuses on modeling microstructural changes in iron-based ferritic alloys under high-energy neutron irradiation relevant to light water fission reactor pressure vessels. We also touch briefly on radiation effects in structural alloys for fusion reactor first wall and blanket structures; in this case the focus is on modeling the evolution of self-interstitial atom clusters and dislocation loops. Note, since even the narrower topic of structural materials for nuclear energy applications encompass a vast literature dating from 1942, the references included in this article are primarily limited to these two narrower subjects. Thus, the references cited here are presented as examples, rather than comprehensive bibliographies. However, the interested reader is referred to proceedings of continuing symposia series that have been sponsored by several organizations, several monographs [2-4] and key journals (e.g., Journal of Nuclear Materials, Radiation Effects and Defects in Solids).

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

Synchronized fusion development considering physics, materials and heat transfer

NASA Astrophysics Data System (ADS)

Wong, C. P. C.; Liu, Y.; Duan, X. R.; Xu, M.; Li, Q.; Feng, K. M.; Zheng, G. Y.; Li, Z. X.; Wang, X. Y.; Li, B.; Zhang, G. S.

2017-12-01

Significant achievements have been made in the last 60 years in the development of fusion energy with the tokamak configuration. Based on the accumulated knowledge, the world is embarking on the construction and operation of ITER (International Thermonuclear Experimental Reactor) with a production of 500 MWf fusion power and the demonstration of physics Q  =  10. ITER will demonstrate D-T burn physics for a duration of a few hundred seconds to prepare for the next long-burn or steady state nuclear testing tokamak operating at much higher neutron fluence. With the evolution into a steady state nuclear device, such as the China Fusion Engineering Test Reactor (CFETR), it is necessary to examine the boundary conditions imposed by the combined development of tokamak physics, fusion materials and fusion technology for a reactor. The development of ferritic steel alloys as the structural material suitable for use at high neutron fluence leads to the use of helium as the most likely reactor coolant. This points to the fundamental technology limitation on the removal of chamber wall maximum heat flux at around 1 MW m-2 and an average heat flux of 0.1 MW m-2 for the next test reactor. Future reactor performance will then depend on the control of spatial and temporal edge heat flux peaking in order to increase the average heat flux to the chamber wall. With these severe material and technological limitations, system studies were used to scope out a few robust steady state synchronized fusion reactor (SFR) designs. As an example, a low fusion power design at 131.6 MWf, which can satisfy steady state design requirements, would have a major radius of 5.5 m and minor radius of 1.6 m. Such a design with even more advanced structural materials like W f/W composite could allow higher performance and provide a net electrical production of 62 MWe. These can be incorporated into the CFETR program.

Using Additive Manufacturing to Optimize FLiBe Coolant Blanket in Fusion Reactors

NASA Astrophysics Data System (ADS)

Fry, Vincent Michael

Fusion reactors have often been hailed as the holy grail of clean energy generation, though a power-generating reactor has never been built due to a multitude of limiting factors. One such factor is the immense 12-15 MW/m2 heat fluxes experienced by the inner wall of the reactor. Multiple groups have proposed the use of tungsten swirl tubes to withstand the heat generated within the reactor core. The primary focus of this investigation is to parameterize this 'first wall' interior structure to determine the highest achievable heat transfer coefficient given the many tungsten configurations enabled via additive manufacturing. Two general tube structures were considered: an orthogonal three-dimensional mesh of various diameters and spacings, as well as a swirl tube geometry with varying 'tape' thicknesses. The coolant liquid proposed is FLiBe (2LiF-BeF2) due to its high specific heat capacity as well as its ability to breed tritium, the fuel for the reactor. This was accomplished using theoretical calculations; computational fluid dynamics and conjugate heat transfer simulations in ANSYS Workbench; as well as an experimental setup to confirm tube pressure drop along the pipe. It was determined that heat transfer coefficients between upwards of 60,000 W/m 2K were readily achievable, keeping the first wall temperature around 1300 K. A multitude of designs proved to be feasible given the pumping power restrictions, though the suggested design going forward is a swirl tube with 2 mm 'tape' thickness and 3 m/s inlet velocity. Simulated pressure drop with water was accurate to within 30% of experimentally measured values, giving confidence in the credibility of the results.

Thermal margin protection system for a nuclear reactor

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

Musick, C.R.

1974-02-12

A thermal margin protection system for a nuclear reactor is described where the coolant flow flow trip point and the calculated thermal margin trip point are switched simultaneously and the thermal limit locus is made more restrictive as the allowable flow rate is decreased. The invention is characterized by calculation of the thermal limit Locus in response to applied signals which accurately represent reactor cold leg temperature and core power; cold leg temperature being corrected for stratification before being utilized and reactor power signals commensurate with power as a function of measured neutron flux and thermal energy added to themore » coolant being auctioneered to select the more conservative measure of power. The invention further comprises the compensation of the selected core power signal for the effects of core radial peaking factor under maximum coolant flow conditions. (Official Oazette)« less

Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

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

Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A.

1997-02-01

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or {open_quotes}recovery,{close_quotes} of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices ofmore » Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed.« less

On the radiation damage characterization of candidate first wall materials in a fusion reactor using various molten salts

NASA Astrophysics Data System (ADS)

Übeyli, Mustafa

2006-12-01

Evaluating radiation damage characteristics of structural materials considered to be used in fusion reactors is very crucial. In fusion reactors, the highest material damage occurs in the first wall because it will be exposed to the highest neutron, gamma ray and charged particle currents produced in the fusion chamber. This damage reduces the lifetime of the first wall material and leads to frequent replacement of this material during the reactor operation period. In order to decrease operational cost of a fusion reactor, lifetime of the first wall material should be extended to reactor's lifetime. Using a protective flowing liquid wall between the plasma and first wall can decrease the radiation damage on first wall and extend its lifetime to the reactor's lifetime. In this study, radiation damage characterization of various low activation materials used as first wall material in a magnetic fusion reactor blanket using a liquid wall was made. Various coolants (Flibe, Flibe + 4% mol ThF 4, Flibe + 8% mol ThF 4, Li 20Sn 80) were used to investigate their effect on the radiation damage of first wall materials. Calculations were carried out by using the code Scale4.3 to solve Boltzmann neutron transport equation. Numerical results brought out that the ferritic steel with Flibe based coolants showed the best performance with respect to radiation damage.

Effects of Welding Parameters on Mechanical Properties in Electron Beam Welded CuCrZr Alloy Plates

NASA Astrophysics Data System (ADS)

Jaypuria, Sanjib; Doshi, Nirav; Pratihar, Dilip Kumar

2018-03-01

CuCrZr alloys are attractive structural materials for plasma-facing components (PFC) and heat sink element in the International Thermonuclear Experimental Reactor (ITER) fusion reactors. This material has gained so much attention because of its high thermal conductivity and fracture toughness, high resistance to radiation damage and stability at elevated temperatures. The objective of this work is to study the effects of electron beam welding parameters on the mechanical strength of the butt welded CuCrZr joint. Taguchi method is used as the design of experiments to optimize the input parameters, such as accelerating voltage, beam current, welding speed, oscillation amplitude and frequency. The joint strength and ductility are the desired responses, which are measured through ultimate tensile strength and percent elongation, respectively. Accelerating voltage and welding speed are found to have significant influence on the strength. A combination of low amplitude and high-frequency oscillation is suggested for the higher joint strength and ductility. There is a close agreement between Taguchi predicted results and experimental ones. Fractographic analysis of joint and weld zone analysis are carried out to study the failure behaviour and microstructural variation in the weld zone, respectively.

Evolution of thermo-physical properties and annealing of fast neutron irradiated boron carbide

NASA Astrophysics Data System (ADS)

Gosset, Dominique; Kryger, Bernard; Bonal, Jean-Pierre; Verdeau, Caroline; Froment, Karine

2018-03-01

Boron carbide is widely used as a neutron absorber in most nuclear reactors, in particular in fast neutron ones. The irradiation leads to a large helium production (up to 1022/cm3) together with a strong decrease of the thermal conductivity. In this paper, we have performed thermal diffusivity measurements and X-ray diffraction analyses on boron carbide samples coming from control rods of the French Phenix LMFBR reactor. The burnups range from 1021 to 8.1021/cm3. We first confirm the strong decrease of the thermal conductivity at the low burnup, together with high microstructural modifications: swelling, large micro-strains, high defects density, and disordered-like material conductivity. We observe the microstructural parameters are highly anisotropic, with high micro-strains and flattened coherent diffracting domains along the (00l) direction of the hexagonal structure. Performing heat treatments up to high temperature (2200 °C) allows us to observe the material thermal conductivity and microstructure restoration. It then appears the thermal conductivity healing is correlated to the micro-strain relaxation. We then assume the defects responsible for most of the damage are the helium bubbles and the associated stress fields.

The microbial community of a biofilm contact reactor for the treatment of winery wastewater.

PubMed

de Beer, D M; Botes, M; Cloete, T E

2018-02-01

To utilize a three-tiered approach to provide insight into the microbial community structure, the spatial distribution and the metabolic capabilities of organisms of a biofilm in the two towers of a high-rate biological contact reactor treating winery wastewater. Next-generation sequencing indicated that bacteria primarily responsible for the removal of carbohydrates, sugars and alcohol were more abundant in tower 1 than tower 2 while nitrifying and denitrifying bacteria were more abundant in tower 2. Yeast populations differed in each tower. Fluorescent in situ hybridization coupled with confocal microscopy showed distribution of organisms confirming an oxygen gradient across the biofilm depth. The Biolog system (ECO plates) specified the different carbon-metabolizing profiles of the two biofilms. The three-tiered approach confirmed that the addition of a second subunit to the bioreactor, expanded the treatment capacity by augmenting the microbial and metabolic diversity of the system, improving the treatment scope of the system. A three-tiered biofilm analysis provided data required to optimize the design of a bioreactor to provide favourable conditions for the development of a microbial consortium, which has optimal waste removal properties for the treatment requirements at hand. © 2017 The Society for Applied Microbiology.

The energy release and temperature field in the ultracold neutron source of the WWR-M reactor at the Petersburg Nuclear Physics Institute

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

Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Kislitsin, B. V.; Onegin, M. S.

2016-12-15

Results of calculations of energy releases and temperature fields in the ultracold neutron source under design at the WWR-M reactor are presented. It is shown that, with the reactor power of 18 MW, the power of energy release in the 40-L volume of the source with superfluid helium will amount to 28.5 W, while 356 W will be released in a liquid-deuterium premoderator. The lead shield between the reactor core and the source reduces the radiative heat release by an order of magnitude. A thermal power of 22 kW is released in it, which is removed by passage of water.more » The distribution of temperatures in all components of the vacuum structure is presented, and the temperature does not exceed 100°C at full reactor power. The calculations performed make it possible to go to design of the source.« less

Implicit time-integration method for simultaneous solution of a coupled non-linear system

NASA Astrophysics Data System (ADS)

Watson, Justin Kyle

Historically large physical problems have been divided into smaller problems based on the physics involved. This is no different in reactor safety analysis. The problem of analyzing a nuclear reactor for design basis accidents is performed by a handful of computer codes each solving a portion of the problem. The reactor thermal hydraulic response to an event is determined using a system code like TRAC RELAP Advanced Computational Engine (TRACE). The core power response to the same accident scenario is determined using a core physics code like Purdue Advanced Core Simulator (PARCS). Containment response to the reactor depressurization in a Loss Of Coolant Accident (LOCA) type event is calculated by a separate code. Sub-channel analysis is performed with yet another computer code. This is just a sample of the computer codes used to solve the overall problems of nuclear reactor design basis accidents. Traditionally each of these codes operates independently from each other using only the global results from one calculation as boundary conditions to another. Industry's drive to uprate power for reactors has motivated analysts to move from a conservative approach to design basis accident towards a best estimate method. To achieve a best estimate calculation efforts have been aimed at coupling the individual physics models to improve the accuracy of the analysis and reduce margins. The current coupling techniques are sequential in nature. During a calculation time-step data is passed between the two codes. The individual codes solve their portion of the calculation and converge to a solution before the calculation is allowed to proceed to the next time-step. This thesis presents a fully implicit method of simultaneous solving the neutron balance equations, heat conduction equations and the constitutive fluid dynamics equations. It discusses the problems involved in coupling different physics phenomena within multi-physics codes and presents a solution to these problems. The thesis also outlines the basic concepts behind the nodal balance equations, heat transfer equations and the thermal hydraulic equations, which will be coupled to form a fully implicit nonlinear system of equations. The coupling of separate physics models to solve a larger problem and improve accuracy and efficiency of a calculation is not a new idea, however implementing them in an implicit manner and solving the system simultaneously is. Also the application to reactor safety codes is new and has not be done with thermal hydraulics and neutronics codes on realistic applications in the past. The coupling technique described in this thesis is applicable to other similar coupled thermal hydraulic and core physics reactor safety codes. This technique is demonstrated using coupled input decks to show that the system is solved correctly and then verified by using two derivative test problems based on international benchmark problems the OECD/NRC Three mile Island (TMI) Main Steam Line Break (MSLB) problem (representative of pressurized water reactor analysis) and the OECD/NRC Peach Bottom (PB) Turbine Trip (TT) benchmark (representative of boiling water reactor analysis).

PRESSURE SYSTEM CONTROL

DOEpatents

Esselman, W.H.; Kaplan, G.M.

1961-06-20

The control of pressure in pressurized liquid systems, especially a pressurized liquid reactor system, may be achieved by providing a bias circuit or loop across a closed loop having a flow restriction means in the form of an orifice, a storage tank, and a pump connected in series. The subject invention is advantageously utilized where control of a reactor can be achieved by response to the temperature and pressure of the primary cooling system.

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.

Recent advances in modeling and simulation of the exposure and response of tungsten to fusion energy conditions

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

Marian, Jaime; Becquart, Charlotte S.; Domain, Christophe

2017-06-09

Under the anticipated operating conditions for demonstration magnetic fusion reactors beyond ITER, structural materials will be exposed to unprecedented conditions of irradiation, heat flux, and temperature. While such extreme environments remain inaccessible experimentally, computational modeling and simulation can provide qualitative and quantitative insights into materials response and complement the available experimental measurements with carefully validated predictions. For plasma facing components such as the first wall and the divertor, tungsten (W) has been selected as the best candidate material due to its superior high-temperature and irradiation properties. In this paper we provide a review of recent efforts in computational modeling ofmore » W both as a plasma-facing material exposed to He deposition as well as a bulk structural material subjected to fast neutron irradiation. We use a multiscale modeling approach –commonly used as the materials modeling paradigm– to define the outline of the paper and highlight recent advances using several classes of techniques and their interconnection. We highlight several of the most salient findings obtained via computational modeling and point out a number of remaining challenges and future research directions« less

Azimuthally anisotropic hydride lens structures in Zircaloy 4 nuclear fuel cladding: High-resolution neutron radiography imaging and BISON finite element analysis

NASA Astrophysics Data System (ADS)

Lin, Jun-Li; Zhong, Weicheng; Bilheux, Hassina Z.; Heuser, Brent J.

2017-12-01

High-resolution neutron radiography has been used to image bulk circumferential hydride lens particles in unirradiated Zircaloy 4 tubing cross section specimens. Zircaloy 4 is a common light water nuclear reactor (LWR) fuel cladding; hydrogen pickup, hydride formation, and the concomitant effect on the mechanical response are important for LWR applications. Ring cross section specimens with three hydrogen concentrations (460, 950, and 2830 parts per million by weight) and an as-received reference specimen were imaged. Azimuthally anisotropic hydride lens particles were observed at 950 and 2830 wppm. The BISON finite element analysis nuclear fuel performance code was used to model the system elastic response induced by hydride volumetric dilatation. The compressive hoop stress within the lens structure becomes azimuthally anisotropic at high hydrogen concentrations or high hydride phase fraction. This compressive stress anisotropy matches the observed lens anisotropy, implicating the effect of stress on hydride formation as the cause of the observed lens azimuthal asymmetry. The cause and effect relation between compressive stress and hydride lens anisotropy represents an indirect validation of a key BISON output, the evolved hoop stress associated with hydride formation.

Structured catalyst bed and method for conversion of feed materials to chemical products and liquid fuels

DOEpatents

Wang, Yong , Liu; Wei, [Richland, WA

2012-01-24

The present invention is a structured monolith reactor and method that provides for controlled Fischer-Tropsch (FT) synthesis. The invention controls mass transport limitations leading to higher CO conversion and lower methane selectivity. Over 95 wt % of the total product liquid hydrocarbons obtained from the monolithic catalyst are in the carbon range of C.sub.5-C.sub.18. The reactor controls readsorption of olefins leading to desired products with a preselected chain length distribution and enhanced overall reaction rate. And, liquid product analysis shows readsorption of olefins is reduced, achieving a narrower FT product distribution.

Structure and activity of a new low-molecular-weight heparin produced by enzymatic ultrafiltration.

PubMed

Fu, Li; Zhang, Fuming; Li, Guoyun; Onishi, Akihiro; Bhaskar, Ujjwal; Sun, Peilong; Linhardt, Robert J

2014-05-01

The standard process for preparing the low-molecular-weight heparin (LMWH) tinzaparin, through the partial enzymatic depolymerization of heparin, results in a reduced yield because of the formation of a high content of undesired disaccharides and tetrasaccharides. An enzymatic ultrafiltration reactor for LMWH preparation was developed to overcome this problem. The behavior, of the heparin oligosaccharides and polysaccharides using various membranes and conditions, was investigated to optimize this reactor. A novel product, LMWH-II, was produced from the controlled depolymerization of heparin using heparin lyase II in this optimized ultrafiltration reactor. Enzymatic ultrafiltration provides easy control and high yields (>80%) of LMWH-II. The molecular weight properties of LMWH-II were similar to other commercial LMWHs. The structure of LMWH-II closely matched heparin's core structural features. Most of the common process artifacts, present in many commercial LWMHs, were eliminated as demonstrated by 1D and 2D nuclear magnetic resonance spectroscopy. The antithrombin III and platelet factor-4 binding affinity of LMWH-II were comparable to commercial LMWHs, as was its in vitro anticoagulant activity. © 2014 Wiley Periodicals, Inc. and the American Pharmacists Association.

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

Tan, Lizhen; Yang, Ying; Chen, Tianyi

Advanced nuclear reactors as well as the life extension of light water reactors require advanced alloys capable of satisfactory operation up to neutron damage levels approaching 200 displacements per atom (dpa). Extensive studies, including fundamental theories, have demonstrated the superior resistance to radiation-induced swelling in ferritic steels, primarily inherited from their body-centered cubic (bcc) structure. This study aims at developing nanoprecipitates strengthened advanced ferritic alloys for advanced nuclear reactor applications. To be more specific, this study aims at enhancing the amorphization ability of some precipitates, such as Laves phase and other types of intermetallic phases, through smart alloying strategy, andmore » thereby promote the crystalline®amorphous transformation of these precipitates under irradiation.« less

Correlative Microscopy of Neutron-Irradiated Materials

DOE PAGES

Briggs, Samuel A.; Sridharan, Kumar; Field, Kevin G.

2016-12-31

A nuclear reactor core is a highly demanding environment that presents several unique challenges for materials performance. Materials in modern light water reactor (LWR) cores must survive several decades in high-temperature (300-350°C) aqueous corrosion conditions while being subject to large amounts of high-energy neutron irradiation. Next-generation reactor designs seek to use more corrosive coolants (e.g., molten salts) and even greater temperatures and neutron doses. The high amounts of disorder and unique crystallographic defects and microchemical segregation effects induced by radiation inevitably lead to property degradation of materials. Thus, maintaining structural integrity and safety margins over the course of the reactor'smore » service life thus necessitates the ability to understand and predict these degradation phenomena in order to develop new, radiation-tolerant materials that can maintain the required performance in these extreme conditions.« less

In Situ Monitoring of Ni-based Catalysts during the Synthesis of Propylene Carbonate

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

Ramin, Michael; Reimann, Sven; Grunwaldt, Jan-Dierk

2007-02-02

Three different nickel complexes were catalytically tested in the synthesis of propylene carbonate by carbon dioxide insertion. XAS measurements of the as prepared catalysts confirmed the differences in the structure which led to the varying catalytic activity. The structure of one of the active nickel-based catalysts was followed in situ by X-ray absorption spectroscopy using a specially designed batch reactor cell. The novel batch reactor allows in situ studies in dense carbon dioxide at elevated temperature and high pressure (up to 200 bar) even at the low energy of the nickel K-edge. Hence, important information on the fate of themore » ligands and structural changes under reaction conditions could be gained providing new insight into the reaction mechanism.« less

REACTOR MODERATOR STRUCTURE

DOEpatents

Fraas, A.P.; Tudor, J.J.

1963-08-01

An improved moderator structure for nuclear reactors consists of moderator blocks arranged in horizontal layers to form a multiplicity of vertically stacked columns of blocks. The blocks in each vertical column are keyed together, and a ceramic grid is disposed between each horizontal layer of blocks. Pressure plates cover- the lateral surface of the moderator structure in abutting relationship with the peripheral terminal lengths of the ceramic grids. Tubular springs are disposed between the pressure plates and a rigid external support. The tubular springs have their axes vertically disposed to facilitate passage of coolant gas through the springs and are spaced apart a selected distance such that at sonae preselected point of spring deflection, the sides of the springs will contact adjacent springs thereby causing a large increase in resistance to further spring deflection. (AEC)

Pulsed eddy current inspection of broach support plates in steam generators

NASA Astrophysics Data System (ADS)

Mokros, Sarah Gwendolyn

Steam Generators (SGs) are a critical component of nuclear reactors, employing thousands of SG tubes to convert heat generated in the reactor core into useable energy. SG tubes are supported at numerous locations by Broach Support Plates (BSPs) that have trefoil shaped holes, which prevent excessive tube vibrations, while allowing water to easily flow through the support structures. A number of degradation modes occur in SGs, such as SG tube fretting, cracking or denting, requiring periodic inspection. Currently, conventional Eddy Current Testing (ECT) is used to non-destructively assess the condition of SG tubes and components. However, as reactors age, new modes of degradation will likely appear that may be difficult to detect and characterize using conventional ECT, such as wall loss in BSPs and build-up of corrosion products, which typically form as a hard sludge called magnetite. Pulsed Eddy Current (PEC) technologies are an emerging technique that is presented in this work as a method to further advance inspection techniques used in CANDURTM nuclear reactors. A PEC probe was designed to inspect the unique shape of the trefoil shaped hole to detect and characterize wall loss and the presence of magnetite in A516 carbon steel BSPs with trefoil shaped holes from within 15.9 mm (5/8") Alloy-800 SG tubes. PEC was also used to observe how measurements of wall loss were affected by the presence of magnetite. This work presents Finite Element Method (FEM) simulations and experimental results collected to observe these degradation modes. The probe was demonstrated to be capable of detecting far side wall loss as low as 20%, locating and characterizing the relative permeability of magnetite, and of detecting wall loss when magnetite was present. FEM simulations and experimental results were found to be in good agreement, suggesting that additional investigations of the effects of BSP degradation on PEC signal response may also be performed using FEM models.

Final summary report for 1989 inservice inspection (ISI) of SRS (Savannah River Site) 100-P Reactor tank

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

Morrison, J.M.; Loibl, M.W.

1989-12-15

The integrity of the SRS reactor tanks is a key factor affecting their suitability for continued service since, unlike the external piping system and components, the tanks are virtually irreplaceable. Cracking in various areas of the process water piping systems has occurred beginning in 1960 as a result of several degradation mechanisms, chiefly intergranular stress corrosion cracking (IGSCC) and chloride-induced transgranular cracking. IGSCC, currently the primary degradation mechanism, also occurred in the knuckle'' region (tank wall-to-bottom tube sheet transition piece) unique to C Reactor and was eventually responsible for that reactor being deactivated in 1985. A program of visual examinationsmore » of the SRS reactor tanks was initiated in 1968, which used a specially designed immersible periscope. Under that program the condition of the accessible tank welds and associated heat affected zones (HAZ) was evaluated on a five-year frequency. Prior to 1986, the scope of these inspections comprised approximately 20 percent of the accessible weld area. In late 1986 and early 1987 the scope of the inspections was expanded and a 100 percent visual inspection of accessible welds was performed of the P-, L-, and K-Reactor tanks. Supplemental dye penetrant examinations were performed in L Reactor on selected areas which showed visual indications. No evidence of cracking was detected in any of these inspections of the P-, L-, and K-Reactor tanks. 17 refs., 7 figs.« less

Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.

The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and implementation of a virtual reactivity feedback control loop; testing and thermal hydraulic characterization of the coolant flow paths for two space reactor concepts; and analysis of heat pipe operation during start-up and steady state operation.

Characterization of bacterial and archaeal communities in air-cathode microbial fuel cells, open circuit and sealed-off reactors.

PubMed

Shehab, Noura; Li, Dong; Amy, Gary L; Logan, Bruce E; Saikaly, Pascal E

2013-11-01

A large percentage of organic fuel consumed in a microbial fuel cell (MFC) is lost as a result of oxygen transfer through the cathode. In order to understand how this oxygen transfer affects the microbial community structure, reactors were operated in duplicate using three configurations: closed circuit (CC; with current generation), open circuit (OC; no current generation), and sealed off cathodes (SO; no current, with a solid plate placed across the cathode). Most (98 %) of the chemical oxygen demand (COD) was removed during power production in the CC reactor (maximum of 640 ± 10 mW/m(2)), with a low percent of substrate converted to current (coulombic efficiency of 26.5 ± 2.1 %). Sealing the cathode reduced COD removal to 7 %, but with an open cathode, there was nearly as much COD removal by the OC reactor (94.5 %) as the CC reactor. Oxygen transfer into the reactor substantially affected the composition of the microbial communities. Based on analysis of the biofilms using 16S rRNA gene pyrosequencing, microbes most similar to Geobacter were predominant on the anodes in the CC MFC (72 % of sequences), but the most abundant bacteria were Azoarcus (42 to 47 %) in the OC reactor, and Dechloromonas (17 %) in the SO reactor. Hydrogenotrophic methanogens were most predominant, with sequences most similar to Methanobacterium in the CC and SO reactor, and Methanocorpusculum in the OC reactors. These results show that oxygen leakage through the cathode substantially alters the bacterial anode communities, and that hydrogenotrophic methanogens predominate despite high concentrations of acetate. The predominant methanogens in the CC reactor most closely resembled those in the SO reactor, demonstrating that oxygen leakage alters methanogenic as well as general bacterial communities.

Effect of extended famine conditions on aerobic granular sludge stability in the treatment of brewery wastewater.

PubMed

Corsino, Santo Fabio; di Biase, Alessandro; Devlin, Tanner Ryan; Munz, Giulio; Torregrossa, Michele; Oleszkiewicz, Jan A

2017-02-01

Results obtained from three aerobic granular sludge reactors treating brewery wastewater are presented. Reactors were operated for 60d days in each of the two periods under different cycle duration: (Period I) short 6h cycle, and (Period II) long 12h cycle. Organic loading rates (OLR) varying from 0.7kgCODm -3 d -1 to 4.1kgCODm -3 d -1 were tested. During Period I, granules successfully developed in all reactors, however, results revealed that the feast and famine periods were not balanced and the granular structure deteriorated and became irregular. During Period II at decreased 12h cycle time, granules were observed to develop again with superior structural stability compared to the short 6h cycle time, suggesting that a longer starvation phase enhanced production of proteinaceous EPS. Overall, the extended famine conditions encouraged granule stability, likely because long starvation period favours bacteria capable of storage of energy compounds. Copyright © 2016 Elsevier Ltd. All rights reserved.

Developmental assessment of the Fort St. Vrain version of the Composite HTGR Analysis Program (CHAP-2)

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

Stroh, K.R.

1980-01-01

The Composite HTGR Analysis Program (CHAP) consists of a model-independent systems analysis mainframe named LASAN and model-dependent linked code modules, each representing a component, subsystem, or phenomenon of an HTGR plant. The Fort St. Vrain (FSV) version (CHAP-2) includes 21 coded modules that model the neutron kinetics and thermal response of the core; the thermal-hydraulics of the reactor primary coolant system, secondary steam supply system, and balance-of-plant; the actions of the control system and plant protection system; the response of the reactor building; and the relative hazard resulting from fuel particle failure. FSV steady-state and transient plant data are beingmore » used to partially verify the component modeling and dynamic smulation techniques used to predict plant response to postulated accident sequences.« 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

Project Luna Succendo: The Lunar Evolutionary Growth-Optimized (LEGO) Reactor

NASA Astrophysics Data System (ADS)

Bess, John Darrell

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched within lunar shipments from the Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides 5 kWe using a free-piston Stirling space converter. The overall envelope for a single unit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. The subunits can be placed with centerline distances of approximately 0.6 m in a hexagonal-lattice pattern to provide sufficient neutronic coupling while allowing room for heat rejection and interstitial control. A lattice of six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network Future improvements include advances in reactor control methods, fuel form and matrix, determination of shielding requirements, as well as power conversion and heat rejection techniques to generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces such as Mars, other moons, and asteroids.

Radiation resistant fiber Bragg grating in random air-line fibers for sensing applications in nuclear reactor cores.

PubMed

Zaghloul, Mohamed A S; Wang, Mohan; Huang, Sheng; Hnatovsky, Cyril; Grobnic, Dan; Mihailov, Stephen; Li, Ming-Jun; Carpenter, David; Hu, Lin-Wen; Daw, Joshua; Laffont, Guillaume; Nehr, Simon; Chen, Kevin P

2018-04-30

This paper reports the testing results of radiation resistant fiber Bragg grating (FBG) in random air-line (RAL) fibers in comparison with FBGs in other radiation-hardened fibers. FBGs in RAL fibers were fabricated by 80 fs ultrafast laser pulse using a phase mask approach. The fiber Bragg gratings tests were carried out in the core region of a 6 MW MIT research reactor (MITR) at a steady temperature above 600°C and an average fast neutron (>1 MeV) flux >1.2 × 10 14 n/cm 2 /s. Fifty five-day tests of FBG sensors showed less than 5 dB reduction in FBG peak strength after over 1 × 10 20 n/cm 2 of accumulated fast neutron dose. The radiation-induced compaction of FBG sensors produced less than 5.5 nm FBG wavelength shift toward shorter wavelength. To test temporal responses of FBG sensors, a number of reactor anomaly events were artificially created to abruptly change reactor power, temperature, and neutron flux over short periods of time. The thermal sensitivity and temporal responses of FBGs were determined at different accumulated doses of neutron flux. Results presented in this paper reveal that temperature-stable Type-II FBGs fabricated in radiation-hardened fibers can survive harsh in-pile conditions. Despite large parameter drift induced by strong nuclear radiation, further engineering and innovation on both optical fibers and fiber devices could lead to useful fiber sensors for various in-pile measurements to improve safety and efficiency of existing and next generation nuclear reactors.

Technicians Manufacture a Nozzle for the Kiwi B-1-B Engine

NASA Image and Video Library

1964-05-21

Technicians manufacture a nozzle for the Kiwi B-1-B nuclear rocket engine in the Fabrication Shop’s vacuum oven at the National Aeronautics and Space Administration (NASA) Lewis Research Center. The Nuclear Engine for Rocket Vehicle Applications (NERVA) was a joint NASA and Atomic Energy Commission (AEC) endeavor to develop a nuclear-powered rocket for both long-range missions to Mars and as a possible upper-stage for the Apollo Program. The early portion of the program consisted of basic reactor and fuel system research. This was followed by a series of Kiwi reactors built to test basic nuclear rocket principles in a non-flying nuclear engine. The next phase, NERVA, would create an entire flyable engine. The final phase of the program, called Reactor-In-Flight-Test, would be an actual launch test. The AEC was responsible for designing the nuclear reactor and overall engine. NASA Lewis was responsible for developing the liquid-hydrogen fuel system. The turbopump, which pumped the fuels from the storage tanks to the engine, was the primary tool for restarting the engine. The NERVA had to be able to restart in space on its own using a safe preprogrammed startup system. Lewis researchers endeavored to design and test this system. This non-nuclear Kiwi engine, seen here, was being prepared for tests at Lewis’ High Energy Rocket Engine Research Facility (B-1) located at Plum Brook Station. The tests were designed to start an unfueled Kiwi B-1-B reactor and its Aerojet Mark IX turbopump without any external power.

EVALUATION OF THE DURABILITY OF THE STRUCTURAL CONCRETE OF REACTOR BUILDINGS AT SRS

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

Duncan, A.; Reigel, M.

2011-02-28

The Department of Energy (DOE) intends to close 100-150 facilities in the DOE complex using an in situ decommissioning (ISD) strategy that calls for grouting the below-grade interior volume of the structure and leaving the above-grade interior open or demolishing it and disposing of it in the slit trenches in E Area. These closures are expected to persist and remain stable for centuries, but there are neither facility-specific monitoring approaches nor studies on the rate of deterioration of the materials used in the original construction or on the ISD components added during closure (caps, sloped roofs, etc). This report willmore » focus on the evaluation of the actual aging/degradation of the materials of construction used in the ISD structures at Savannah River Site (SRS) above grade, specifically P & R reactor buildings. Concrete blocks (six 2 to 5 ton blocks) removed from the outer wall of the P Reactor Building were turned over to SRNL as the first source for concrete cores. Larger cores were received as a result of grouting activities in P and R reactor facilities. The cores were sectioned and evaluated using microscopy, x-ray diffraction (XRD), ion chromatography (IC) and thermal analysis. Scanning electron microscopy shows that the aggregate and cement phases present in the concrete are consistent with the mix design and no degradation mechanisms are evident at the aggregate-cement interfaces. Samples of the cores were digested and analyzed for chloride ingress as well as sulfate attack. The concentrations of chloride and sulfate ions did not exceed the limits of the mix design and there is no indication of any degradation due to these mechanisms. Thermal analysis on samples taken along the longitudinal axis of the cores show that there is a 1 inch carbonation layer (i.e., no portlandite) present in the interior wall of the reactor building and a negligible carbonation layer in the exterior wall. A mixed layer of carbonate and portlandite extends deeper into the interior (2-3 inches) and exterior (1-2 inches) walls. This is more extensive than measured in previous SRS structures. Once the completely carbonated layer reaches the rebar that is approximately 2-3 inches into the concrete wall, the steel is susceptible to corrosion. The growth rate of the carbonated layer was estimated from current observations and previous studies. Based on the estimated carbonation rate, the steel rebar should be protected from carbonation induced corrosion for at least another 100 years. If degradation of these structures is dominated by the carbonation mechanism, the length of time before water intrusion is expected into the process room of P-reactor is estimated to be between 425-675 years.« less

AGC 2 Irradiated Material Properties Analysis

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

Rohrbaugh, David Thomas

2017-05-01

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core componentsmore » within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.« less

AGC 2 Irradiation Creep Strain Data Analysis

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

Windes, William E.; Rohrbaugh, David T.; Swank, W. David

2016-08-01

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. Nuclear graphite H-451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within amore » commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.« less

Micro-structural study and Rietveld analysis of fast reactor fuels: U-Mo fuels

NASA Astrophysics Data System (ADS)

Chakraborty, S.; Choudhuri, G.; Banerjee, J.; Agarwal, Renu; Khan, K. B.; Kumar, Arun

2015-12-01

U-Mo alloys are the candidate fuels for both research reactors and fast breeder reactors. In-reactor performance of the fuel depends on the microstructural stability and thermal properties of the fuel. To improve the fuel performance, alloying elements viz. Zr, Mo, Nb, Ti and fissium are added in the fuel. The first reactor fuels are normally prepared by injection casting. The objective of this work is to compare microstructure, phase-fields and hardness of as-cast four different U-Mo alloy (2, 5, 10 and 33 at.% Mo) fuels with the equilibrium microstructure of the alloys. Scanning electron microscope with energy dispersive spectrometer and optical microscope have been used to characterize the morphology of the as-cast and annealed alloys. The monoclinic α'' phase in as-cast U-10 at.% Mo alloy has been characterized through Rietveld analysis. A comparison of metallographic and Rietveld analysis of as-cast (dendritic microstructure) and annealed U-33 at.% Mo alloy, corresponding to intermetallic compound, has been reported here for the first time. This study will provide in depth understanding of microstructural and phase evolution of U-Mo alloys as fast reactor fuel.

Synthesis of colloidal metal nanocrystals in droplet reactors: the pros and cons of interfacial adsorption.

PubMed

Zhang, Lei; Wang, Yi; Tong, Limin; Xia, Younan

2014-07-09

Droplet reactors have received considerable attention in recent years as an alternative route to the synthesis and potentially high-volume production of colloidal metal nanocrystals. Interfacial adsorption will immediately become an important issue to address when one seeks to translate a nanocrystal synthesis from batch reactors to droplet reactors due to the involvement of higher surface-to-volume ratios for the droplets and the fact that nanocrystals tend to be concentrated at the water-oil interface. Here we report a systematic study to compare the pros and cons of interfacial adsorption of metal nanocrystals during their synthesis in droplet reactors. On the one hand, interfacial adsorption can be used to generate nanocrystals with asymmetric shapes or structures, including one-sixth-truncated Ag octahedra and Au-Ag nanocups. On the other hand, interfacial adsorption has to be mitigated to obtain nanocrystals with uniform sizes and controlled shapes. We confirmed that Triton X-100, a nonionic surfactant, could effectively alleviate interfacial adsorption while imposing no impact on the capping agent typically needed for a shape-controlled synthesis. With the introduction of a proper surfactant, droplet reactors offer an attractive platform for the continuous production of colloidal metal nanocrystals.

Design of conduction cooling system for a high current HTS DC reactor

NASA Astrophysics Data System (ADS)

Dao, Van Quan; Kim, Taekue; Le Tat, Thang; Sung, Haejin; Choi, Jongho; Kim, Kwangmin; Hwang, Chul-Sang; Park, Minwon; Yu, In-Keun

2017-07-01

A DC reactor using a high temperature superconducting (HTS) magnet reduces the reactor’s size, weight, flux leakage, and electrical losses. An HTS magnet needs cryogenic cooling to achieve and maintain its superconducting state. There are two methods for doing this: one is pool boiling and the other is conduction cooling. The conduction cooling method is more effective than the pool boiling method in terms of smaller size and lighter weight. This paper discusses a design of conduction cooling system for a high current, high temperature superconducting DC reactor. Dimensions of the conduction cooling system parts including HTS magnets, bobbin structures, current leads, support bars, and thermal exchangers were calculated and drawn using a 3D CAD program. A finite element method model was built for determining the optimal design parameters and analyzing the thermo-mechanical characteristics. The operating current and inductance of the reactor magnet were 1,500 A, 400 mH, respectively. The thermal load of the HTS DC reactor was analyzed for determining the cooling capacity of the cryo-cooler. The study results can be effectively utilized for the design and fabrication of a commercial HTS DC reactor.

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

Steady-state generation of hydrogen peroxide: kinetics and stability of alcohol oxidase immobilized on nanoporous alumina.

PubMed

Kjellander, Marcus; Götz, Kathrin; Liljeruhm, Josefine; Boman, Mats; Johansson, Gunnar

2013-04-01

Alcohol oxidase from Pichia pastoris was immobilized on nanoporous aluminium oxide membranes by silanization and activation by carbonyldiimidazole to create a flow-through enzyme reactor. Kinetic analysis of the hydrogen peroxide generation was carried out for a number of alcohols using a subsequent reaction with horseradish peroxidase and ABTS. The activity data for the immobilized enzyme showed a general similarity with literature data in solution, and the reactor could generate 80 mmol H2O2/h per litre reactor volume. Horseradish peroxidase was immobilized by the same technique to construct bienzymatic modular reactors. These were used in both single pass mode and circulating mode. Pulsed injections of methanol resulted in a linear relation between response and concentration, allowing quantitative concentration measurement. The immobilized alcohol oxidase retained 58 % of initial activity after 3 weeks of storage and repeated use.

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

Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

The PIUS Advanced Reactor is a 640-MW(e) pressurized-water reactor developed by Asea Brown Boveri. A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity normally is controlled by the boron concentration in the coolant and the temperature of the moderator coolant. Analyses of five initiating events have been completed on the basis of calculations performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. The initiating events analyzed are (1) reactor scram, (2) loss of off-site power (3) main steam-line break, (4) small-break loss of coolant, and (5) large-break loss of coolant. Inmore » addition to the baseline calculation for each sequence, sensitivity studies were performed to explore the response of the PIUS reactor to severe off-normal conditions having a very low probability of occurrence. The sensitivity studies provide insights into the robustness of the design.« 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.

Molecular assessment of microbiota structure and dynamics along mixed olive oil and winery wastewaters biotreatment.

PubMed

Eusébio, Ana; Tacão, Marta; Chaves, Sandra; Tenreiro, Rogério; Almeida-Vara, Elsa

2011-07-01

The major parcel of the degradation occurring along wastewater biotreatments is performed either by the native microbiota or by added microbial inocula. The main aim of this study was to apply two fingerprinting methods, temperature gradient gel electrophoresis (TGGE) and length heterogeneity-PCR (LH-PCR) analysis of 16S rRNA gene fragments, in order to assess the microbiota structure and dynamics during mixed olive oil and winery wastewaters aerobic biotreatment performed in a jet-loop reactor (JLR). Sequence homology analysis showed the presence of bacterial genera Gluconacetobacter, Klebsiella, Lactobacillus, Novosphingobium, Pseudomonas, Prevotella, Ralstonia, Sphingobium and Sphingomonas affiliated with five main phylogenetic groups: alpha-, beta- and gamma-Proteobacteria, Firmicutes and Bacteroidetes. LH-PCR analysis distinguished eight predominant DNA fragments correlated with the samples showing highest performance (COD removal rates of 67 up to 75%). Cluster analysis of both TGGE and LH-PCR fingerprinting profiles established five main clusters, with similarity coefficients higher than 79% (TGGE) and 62% (LH-PCR), and related with hydraulic retention time, indicating that this was the main factor responsible for the shifts in the microbiota structure. Canonical correspondence analysis revealed that changes observed on temperature and O(2) level were also responsible for shifts in microbiota composition. Community level metabolic profile analysis was used to test metabolic activities in samples. Integrated data revealed that the microbiota structure corresponds to bacterial groups with high degradative potential and good suitability for this type of effluents biotreatments.

ETR, TRA642. ON GROUND FLOOR, CAMERA LOOKS SOUTHWEST INTO PIT. ...

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

ETR, TRA-642. ON GROUND FLOOR, CAMERA LOOKS SOUTHWEST INTO PIT. CANAL STRUCTURE IS AT RIGHT OF CENTER WITH RECTANGULAR OPENING TO BE MATED WITH THE DE-FUELING MECHANISM THAT WILL DEPOSIT FUEL RODS INTO THE WORKING CANAL. INL NEGATIVE NO. 56-3710. R.G. Larsen, Photographer, 11/13/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Possible criticality of marine reactors dumped in the Kara Sea

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

Warden, J.M.; Mount, M.; Lynn, N.M.

1997-05-01

The largest inventory of radioactive materials dumped in the Kara Sea by the former Soviet Union comes from the spent nuclear fuel (SNF) of seven marine reactors. Using corrosion models derived for the International Arctic Seas Assessment Project (IASAP), the possibility of some of the SNF achieving criticality through structural and material changes has been investigated. Although remote, the possibility cannot at this stage be ruled out.

Tritium

DTIC Science & Technology

2011-11-01

fusion energy -production processes of the particular type of reactor using a lithium (Li) blanket or related alloys such as the Pb-17Li eutectic. As such, tritium breeding is intimately connected with energy production, thermal management, radioactivity management, materials properties, and mechanical structures of any plausible future large-scale fusion power reactor. JASON is asked to examine the current state of scientific knowledge and engineering practice on the physical and chemical bases for large-scale tritium